ML20213F835

Forwards Safety Evaluation Justifying Use of Pressurizer pressure-high Reactor Trip Setpoint of 2,375 Psig & Supporting Required Tech Spec Change Concern Re Calculation of ESF Response Time for Safety Injection Function

ML20213F835
Person / Time
Site:	Beaver Valley
Issue date:	05/07/1987
From:	Carey J DUQUESNE LIGHT CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
2NRC-7-108, TAC-62942, NUDOCS 8705180008
Download: ML20213F835 (19)

Text

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USSne @ 2NRC-7-108 (412)393-7646 Beaver Valley No 2 Unit Project organuation Telecopy(412)393 7889 Sto,"="a May 7, 1987 Shippingport. PA 15077 United States Nuclear Regulatory Comnission ATTN: Document Control Desk Washington, DC 20555

SUBJECT:

Beaver Valley Power Station Unit No. 2 Docket No. 50-412 Tech nical Specification Additional Information Gentlemen:

The Staff has raised several questions during their review of the Beaver Valley Unit 2 Technical Specifications. Attachment A to this letter provides a safety evaluation justification for the use of a pressurtzer pressure-high reactor trip setpoint of 23/5 psig as indicated in the " Westinghouse Setpoint Methodology for Protection Systems, Beaver Valloy Unit 2" (WCAP-11366/11367) transmitted previously.

Attachment R explains the apparent discrepancy raised by the Staff between the FSAR Chapter 15 LOCA analysts assumption for $1 response timo and the Tech-nical Specification ESF response time table.

Attachment C provides a safety evaluation supporting the required Technt-cal Specification change as a result of a recent concern in the calculation of ESF response time for the Safety injection functton as assumed by Westinghouse in their FSAR accident analysis performed for Deaver Valley Unit 2. This addresses a generic concern by Westinghouse regarding the sequencing of the RWST and VCT isolation valves following a Safety injection signal. Future FSAR ,

revision to reflect this revised ESF response time is also attached, l DUQUESNE LIGliT COMPANY hibt

~

37ey [b Senior Vice President HWF/tjr NR/RWF/IECH/SPE Attachment l

AR/NAR cci Mr. P. Tam, Project Manaqcr (w/a)

Mr.il.I! call,NRCSr.ResidentInspector(w/a)

Mr. L. Prtvidy, NRC Resident inspector (w/a)

INPO Hecords Center (w/a) l 0700100000 070507 POR ADOCK 05000412 j l p PDH

.o l

ld ATTACHNENT A i

BACKGROUNO The High Pressurizer Pressure Reactor Trip function is credited in the Loss of Load / Turbine Trip analyses for teaver Valley Unit 2 (SVPS-2 FSAR 15.2.2 and 15.2.3). A safety analysis setpoint value of 2410 psig was assumed in the  :

analysis of these events. The statistical setpoint study for Beaver Valley Unit 2 (WCAP-11366/11367), however, determined that in order to maintain the current Technical specification setpoint of 2375 psig for High Pressurizer l Pressure Reactor Trip. a safety analysis limit of 2425 is required. This '

evaluation addresses the safety impact on Beaver Valley Unit 2 for raising the safety analysis limit from 2410 psig to 2425 psig.

EVALUATION  !

A safety evaluation in the form of answers to key safety questions is provided !

below:  !

! 1. ll111 the nrobability of an accident nreviousiv evaluated in the FSAR be j

ncreased?

l No. The proposed increase in the safety analysis setpoint value will have l no impact on the operation of the plant and will have no impact on the f probability of an accident previously evaluated in the FSAR.

2. ll111 the consecuences of an accident areviously evaluated in the FSAR be L nereased? l No. The High Pressurizer Pressure Reactor Trip function is only credited  !

in the Loss of Load / Turbine Trip transients. An evaluation of the impact t of increasing the High Pressurizer Pressure Reactor Trip setpoint from  !

2410 psig to 2425 ps'g has been performed. The results of this evaluation are presented below.

The BVPS-2 FSAR presents the results of four LOL/TT cases for N loop i operation. These four cases are:

Case 1 - Minimum reactivity feedback, without pressurizer PORV's. I Case 2 - Maximum reactivity feedback, without pressurizer P0RV's.

Case 3 - Minimum reactivity feedback, with pressurizer PORV's. '

Case 4 - Maximum reactivity feedback, with pressurizer PORV's.

I Cases 3 and 4 do not trip on High Pressurizer Pressure, and as such are l unaffected by an increase in the High Pressuriter Pressure Reactor Trip '

setpoint. High Pressurizer Pressure reactor trip signals for Cases I and 2 are generated at 5.97 and 5.92 seconds, respectively.

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e a

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!' Attachment A Page 2.

{

Control rod motion is modeled to occur 2 seconds later. A 15 psi increase in the High Pressurizer Pressure reactor trip setpoint from 2410 psig tc.  !

2425 psig would delay reactor trip by approximately 0.2 seconds. This r delay in reactor trip will result in a 6 psi increase in pressurizer l pressure at the time of rod motion. Since the rate of pressure increase prior to rod motion is decreasing and because the pressurizer safety valve relief rate increases with pressure, the increase in the peak pressurizer l pressure due to the increase in the trip setpoint will be less than 6 psi. An increase in pressure of this magnitude will not result in an overpressurization of the RCS. Additionally, the setpoint increase will  ;

not have an adverse impact on the calculated minimum DNSR for the event (DNSR never falls below its initial value throughout the event). Thus the results and conclusions of the FSAR for the Loss of Load / Turbine Trip '

event remain valid for a 15 pst increase in the High Pressurizer Pressure safety analysis limit. 4

3. Nav the nonsibility of an accident which is different than any alreadv i evaluated ' n the F5AR be created?

No. The proposed increase in the safety analysis setpoint value will have no impact on the operation of the plant and will not result in the creation of the possibility of an accident which is different than any  :

already evaluated in the FSAR. '

i

4. Will the erobability of a ma9 function of eauinment imoortant to safety areviousiv eva' usted < n the f5AR be increased?

l I

No. The proposed increase in the safety analysis setpoint value will have no impact on the operation of the plant and will not result in an increase in the probability of a malfunction of equipment important to safety i previously evaluated in the FSAR. '

5. Will the consecuences of a malfunction of eat'oment imoortant to safety I areviously eva' usted in the FSAR be ' nc reasec P 1 No. The proposed increase in the safety analysis setpoint value will have no impact on the operation of the plant and will not result in an increase  :

in the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR.

6. Nav the nossibility of a malfunctior of e nuinmer t immortant to safety difforent than anv a' readv evaluatec '

nt te FSAF be created?

l No. The proposed increase in the safety analysis setpoint value will have no impact on the operation of the plant and will not create the possibility of a malfunction of equipment important to safety difforent than any already evaluated in the FSAR.

i

Page 3.

Attachment A .

7. Will the marain of-safety as defined in the bases to any technical specification be reduced?

No. The proposed increase in the safety analysis setpoint will not reduce the margin of safety as defined in the bases to the Beaver Valley Unit 2 Technical Specifications. -

8. What FSAR changes are needed?

Table 15.0-4 will be revised in a future amendment to reference the safety evaluation in this letter which justifies the use of a trip setpoint of 2425 psig for high pressurizer pressure.

i l

i ATTACHMENT B Ouring the Technical Specification review, the NRC Staff questioned the apparent discrepancy between the FSAR Chapter 15 LOCA analysis assumption for S1 response time and the Technical Specification ESF response time table (Table 3.3-5).

I This letter documents the consistency between the 25 seconds (i.e., 26.420 - j 1.420 seconds = 25.0 seconds as specified in the BVPS-2 FSAR Table 15.6-1, '

attached) for Safety Injection and the 27 seconds specified in Table 3.3-5 items 2(a) and 3(a) of the 8VPS-2 Technical Specifications (also attached) as follows.

For a LOCA coincident with a loss-of-offsite power (LOOP), the safety

! injection pumps are started sequentially as follows:

1. The diesel generators are started within 2 seconds from the time the safety injection setpoint is reached.

2. 12 seconds after the SI setpoint is reached, the diesel generators are up to speed and the charging pumps are started. Within 17 seconds after the SI setpoint is reached, the charging pumps are delivering full flow.

3. The low head safety injection pumps are started 17 seconds after the SI setpoint is reached and are delivering full flow within 22 seconds, i

15 order to simplify the LOCA analysis modal, Westinghouse determined that an analysis assumption of delaying all SI flow for 25 seconds and then assuming full flow was consistent with the plant design where some SI flow is available at 17 seconds and ramps up to full flow by 27 seconds. The 8VPS-2 Technical Specification $1 response time of 27 seconds is therefore consistent with the LOCA analysis assumption of 25 seconds.

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. ATTACHMENT B Page.2 gyps.2 FS&R TABLE 15.6-1 Tint sEgUENCE OF EVENTS FOR INCIDENTS WIIOt CAUSE l A DECREASE IN REACTOR COOL &NT INVENTORY l

Time (sec)

Event N . Loop N-1 Loop Accident Inadvertent opening Relief valve opens 0.0 0.0 of a pressurizer relief valve Low pressuriser 28.49 24.3 l pressure reactor trip set point reached Rods begin to drop 30.69 26.3 Ninimum DNBR occurs 31.2 26.8 2 Large break LOCA DECLG C = 0.8 Start 0.00 D

Reactor trip signal 0.384 Safety injection signal 1.420 Accumulator injection l l begins 9.230

{  !

End-of-bypass 25.399 I

Ind-of-blowdown 25.399 26.420 1

Pump injection begins 39.566 I

Bottom of core recovery ,

j i i

t Accumulator empty 49.001 0.00  !

DECLG g = 0.6 Start I

Reactor trip signal 0.388 I

safety injection signal 1.620 Accumulator injection g

] 11.500

begins 1 of 5 January 1987 Amen h nt 13 i

,c,,-.,- ,,- - .-,- -.-.- -.. ..-.- - - - ,, - -.. ._,_ _ _,..- , - ,, - . -. ,---- ., - - . , . .

M -2 FSAR ATTACHMENT B Page 3.

TABLE 15.6-1 (Cont)

Time (sec)

N Loop N-1 Loop i Accident i Event Pump injection begins 26.620 .

End-of-bypass 28.549 End-of-blowdown 28.549 Bottom of core recovery 42.485 Accumulator Empty 52.045 Start 0.0 DECLG CD = 0.4 Reactor trip signal 0.394 I safety injection signal 1.980 g

Accumulator injection begins 15.500 .

End-of-bypass 34.070

! l End-of-blowdown 34.070 Pump injection begins 26.980 I

l Bottom of core recovery 48.546 l

Accumulator empty 57.678 l

Large break LOCA (N-1) Active Loop Break Start 0.0 DECLG Cp = 0.4 Reactor trip signal 0.37 Safety injecttion signal 1.56 Accumulator injection begins 12.4 End-of-bypass 23.9 End-of-blowdown 25.8 2 of 5 January 1987 Amendment 13 e- -- - ,,--a -w - + - - - - ~ v-,- ~

ATTACHMENT B TABLE 3 3-5

, Page 4.

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL,AND FUNCTION RESPONSE TIME IN SECONDS

1. Manual

a. Safety Injection (ECCS) Not Applicable Feedwater Isolation Not Applicable Reactor Trip (51) Not Applicable .

Not Applicable )

Containment' Isolation-Phase "A" l

Containment Vent and Purge Isolation Not Applicable Auxiliary Feedwater Pumps. Not Applicable Service Water System Not Applicable

b. Containment Quench Spray Pumps Not Applicable

• Containment Quench Spray Valves Not Applicable

~

Containment Isolation-Phase "B" Not Applicable

c. Containment Isolation-Phase "A" Not Applicable

d. Control Room Ventilation Isolation Not Applicable

2. Containment Pressure-High

a. Safety Injection (ECCS) $ 27.0*

b. Reactor Trip (from SI) $ 2.0

c. Feedwater Isolation 1 7.0(1)

d. Containment Isolation-Phase "A" $ 61.5(4)/115.5(5) 4

e. Auxiliary Feedwater Pumps i 60.0

f. Service Water System 5 72.5(2)/181.5(3) 1

3. Pressurizer Pressure-Low

a. Safety Injection (ECCS) $ 27.0*/12.0#

b. Reactor Trip (from SI) i 2.0

c. Feedwater Isolation 5 7.0(1)

d. Containment Isolation-Phase "A" .i 61.0(4)/115.0 (5)

BEAVER VALLEY - UNIT 2 3/4 3-29

• r-y -w y------r v -

w i ATTACHMENT C SAFETY EVALUATION FOR TECHNICAL SPECIFICATION REVISION OF ESF RESPONSE TIMES FOR BEAVER VALLEY UNIT 2 The purpose of this safety evaluation is to provide the justification to change the Beaver Valley Unit 2 (BV-2) Technical Specification ESF response times and response time notations for Containment Pressure - High-1, Low Pressurizer Pressure and Low Steamline Pressure (Table 3.3-5, Items 2a, 3a, 4

and 4a) to reflect the valve interlock in the safety injection system between the Volume Control Tank (VCT) isolation. valves and the Refueling Water Storage Tank (RWST) isolation valves.

BACKGROUND In the normal configuration of the Chemical and Volume Control System (CVCS),

the charging pumps take suction from the Volume Control Tank. When a Safety Injection ("S") signal is generated from the protection logic, a signal is sent to start the centrifugal charging pumps and to begin opening the Refueling Water Storage Tank isolation valves, in order to align the borated water source for delivery to the RCS. Once the RWST isolation valves have repositioned and are indicated fully open, the isolation valves on the VCT will begin to close. This sequential valve stroke time can be as long-as 25 seconds. Since the VCT is pressurized, it will be the source of the SI flow until the isolation valves are closed. This affects the time assumed at which the borated water in the RWST is available to the suction of the charging pumps.

The current Steamline Break analyses (BVPS-2 FSAR Section 15.1.4 and 15.1.5) which support the current Technical Specifications (Table 3.3-5) assume the following delays for the delivery of borated water to the RCS:

1. "S" signal generation (2 seconds);

2. Diesel start, including time to come up to speed (10 seconds);

3. Valve stroke times and pumps to full speed (10 seconds).

However, this assumes that the VCT and RWST isolation valves stroke simultaneously rather than sequentially. The valve interlock logic increases the delay time for the avnilability of borated water by 15 seconds (conservatively) to 27 seconds with offsite power and 37 seconds without offsite power. The only non-LOCA transient impacted by the increased time delay is the Steamline Break event. No other Chapter 15 transient relies on short-term boration from the RWST to mitigate the event.

EVALUATION Non-LOCA Events:

Based on the current Steamline Break analysis for BV-2 and sensitivities performed for other plants, the additional time delay is acceptable.

Specifically:

,/ .

i-Page 2.

~ ATTACHMENT C

1. The additional delay in the availability of borated' water occurs early in

the Steamline Break transient when RCS pressures are relatively high and SI flowrates are relatively small due to head vs. SI flow characteristics.

2. Previous' sensitivities have shown that delays of this. magnitude result in small changes in the analysis results. A BV-2 specific review of the Steamline Break analysis demonstrated that there is sufficient margin

j. available in the analysis such that the conclusions presented in the FSAR remain valid.

~

From analyses performed for other plants, it has been shown that SI boron

!. concentration reduction has little effect on the Steamline Creak mass / energy l

release analysis inside containment. Since the ariditional time delay is a ,

small perturbation compared to a large change i.1 the available boron

concentration, there will be a negligible impact on the Steamline Break

+

mass / energy release inside containment analysis.

l Sensitivities performed for the Steamline.Breair superteated mass / energy release outside containment analysis show that the re:ults are not sensitive ,

, to large changes in SI flow (Reference, supporting do.umentation for i WCAP-10961, Rev.1) . The additional time delay is a small perturbation .

! compared to a large change in total SI flow, therefore, it is concluded that j the impact on the BV-2 superheated mass / energy' releases outside containment is insignificant.

LOCA Events:

[ Larae Break LOCA - FSAR Chapter 15.6.5 l

! Large break LOCA analyses are performed under the assumption that the

insnediate safety f unction of the safety injection-(SI) system is .to supply water to the Reactor Coolant System (RCS),.whether borated or not. The time

at which water (from either the VCT or the RWST) is available to the suction l l

of the charging pumps (high head SI pumps)' depends on the "S" signal l

generation time and the time delay for the pumps to attain full speed. This time will not be affected by the time delay for closure of the VCT isolation

,' valves, since the valve alignment for the supply of borated water to the RCS .I l

is not. considered in the analyses. .Although the pressure at the charging pump inlet will be higher when the VCT valves are open, the charging pump flow rate

, will not be degraded. Also,-negative reactivity insertion due:to core' voiding-causes the nuclear chain reaction to stop and reduce the core power and decay heat levels without reliance on the injection fluid boron concentration.

- Thus, for those SI actuation signals that are only intended to provide.

protection against a LOCA, the additional delay for injection of borated water

. is not required since boron is only required for maintaining subcriticality in-

. the long term following a LOCA.

1 The current FSAR large break LOCA analysis for Beaver Valley Unit 2 was i performed using.the 1981 ECCS Evaluation Model along with the BART Reflood Model (References 2 and 3). and resulted in a peak clad temperature (PCT) of

! l

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- _ _ .[. _ _ - _ _ _ _._. - - - ;_.__ .- . _ _ _ _ . - _ _._. -. _ _ _ _ . . _ _ - -

. 1 o

. 's ATTACHMENT C Page 3.

2120*F for a double-ended _ cold leg guillotine (DECLG) break with a discharge i coefficient of 0.4. ~From the above discussion it can be concluded that a.

. delay in borated water injection will have no -impact on the FSAR large break LOCA ECCS analysis results for Beaver Valley Unit 2, and the current results will still be valid.

4 Small Break LOCA - FSAR Chapter 15.6.5

Small break LOCA analyses are performed under the assumption that the immediate. safety function of the safety injection (SI) system is to supply l

water to the Reactor Coolant System (RCS), whether borated or not. This is similar to the large break LOCA analyses assumption. In the small break LOCA analyses, shutdown of the reactor core is . achieved by insertion of all but the most reactive of the rod control cluster assemblies, an'd does not rely upon the boron concentration. Therefore, as indicated above, for those SI-actuation signals that are only intended to provide protection against a LOCA, the additional delay for injection of borated water will not affect the small break LOCA analysis results. Boron is only required for maintaining

! subcriticality in the long term following a LOCA.

The current FSAR small break LOCA analysis for Beaver-Valley Unit 2 was performed.using the NOTRUMP Small Break LOCA Evaluation Model (Reference 4),

j- which resulted in the most limiting PCT of 1399'F for a 4 inch DECLG break.

From the above discussion it can be concluded that a delay in borated water injection will have no impact on the FSAR small break LOCA ECCS analysis i results for Beaver Valley Unit 2, and the current results will sttil be valid.

Rod Eiection Mass and Enerav Release for Dose Calculations - FSAR ChaDter 15 j The delay in the borated water injection is expected to have no impact on the Rod Ejection Accident analysis for reasons. discussed under the LOCA analyses.

i Containment Intearity (Short and Lona Term Mass and Enerav Release) - FSAR L Chapter 6.2 i

l The containment analyses are described in the FSAR Sections 6. 2.1. 2, 6. 2.1. 3 l and 6.2.2. These sections consider, respectively, the containment

}- subcompartments, Mass and Energy for postulated LOCAs and containment heat

! removal systems. For containment subcompartment analyses, a delay in the injection of the borated water would have no impact on the calculated results, i since the short duration of the transient (< 3 seconds) does not consider any i

safety injection flow. The long term mass and energy release calculations are performed under the assumption that the safety injection system supply flow to the RCS whether borated or not. This is.similar to the LOCA analyses assumptions, and therefore, a delay in'the injection of the borated water 1 would have no impact on the long term mass and energy' releases calculated for Beaver Valley Unit 2.

I P

I. .- . . - - , , - - - - - . _ , . - , . , - , _ - _ - . - - - - -,, __ - . . . . . - . - . . , _ . . . , - . . . . . - - - - , , . _-.- . -

l e :- . l i l ATTACHMENT C - Page 4.

1 1

Steam Generator Tube Ruoture - FSAR ChaDter 15.6.3 The FSAR Steam Generator Tube Rupture (SGTR) accident for Beaver Valley Unit 2 was performed using the.LOFTRAN Model. The primary to secondary break flow was assumed terminated at 30 minutes after initiation of the SGTR event, and the operator recovery action to cooldown the RCS was not modeled in the 1 analysis. Without the RCS cooldown, sufficient shutdown margin is available.

Therefore, the additional delay for injection of the borated water would have '

no impact on the FSAR SGTR analysis for Beaver Valley Unit 2.

Blowdown Reactor Vessel and Loop Forces - FSAR Chapter 3.9N  ;

The b70wdown hydraulic loads resulting from a loss of coolent accident are considered in Section 3.9N.l.4.3 (Reactor Vessel Loss of Coolant Accident Analysis), and Section 3.9N.2.5 (Dynamic Analysis of Reactor Internals Under Faulted Conditions) of Volume 5 of the Beaver Valley Unit 2 FSAR. The increase in the delay time until the availability of borated water as a result- ,

of the VCT/RWST valve interlock logic will not affect the LOCA blowdown  !

hydraulic loads since the maximum loads are generated within the first few I

seconds after break initiation. For this reason, the ECCS and associated i valve interlock logic are'not considered in the LOCA hydraulic forces modeling r 2 and thus the additional water delivered from the VCT during the switchover to

! suction from the RWST does not affect the results of the LOCA hydraulic forces calculations.

4 j Post LOCA Lona Term Core Coolina: Westinahouse Licensina Position - FSAR Chapter 15.6.5 l

j The Westinghouse licensing position for satisfying the requirements of NCFR f Part 50 Section 50.46 Paragraph (b) Item (5) "Long Term Cooling" is-defined in j WCAP-8339 (Reference 5, pp. 4-22). The Westinghouse commitment is that the reactor will remain shutdown by borated ECCS water residing in the sump post i

LOCA (Reference 6). Since credit for the control rods is not taken for large j break LOCA, the borated ECCS water provided by the accumulators and the RWST i must have a concentration that, when mixed with other sources of borated an'd non-borated water, will result in the reactor core remaining subcritical

! assuming all control rods out.

In the normal configuration of the Chemical and Volume Control System (CVCS),

the charging pumps take suction from the VCT. When a Safety Injection (S)

! signal is generated from-the protection logic ~, a signal is sent to start the i centrifgugal charging pumps (Ch/SI) and to begin opening the Refueling Water Storage Tank (RWST) isolation valves in order to align the borated water source for delivery to the RCS. Once the RWST isolation valves have been repositioned and are indicated to be. fully open, the isolation valves on the VCT will begin to close. This sequential. valve stroke time can be as long as

] 25 seconds. Since the VCT is pressurized, it serves as a source of safety injection flow until the isolation valves are closed. Thus, the time assumed at which borated water in the RWST is.available to the suction of the charging pumps is affected, and more importantly for this evaluation, the amount of.

non-borated VCT water that has.been injected into the RCS must be considered.

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. . __ . . ~ -. . _ _ _ _ . . _ _ _ _ _ _ -

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ATTACHMENT C - Page 5.

l~ The ef fect of injecting additional non-borated water into the'RCS during the i

switchover to suction from the RWST has been evaluated with respect to the

• long term cooling - Boron evaluation. It has been concluded that the amount of additional non-bcrated water does not significantly reduce the sump Boron average concentration and therefore does not affect the ability of the core to

~

remqin shutdown by borated'ECCS water. The additional delay time in the

- avaitability of borated water, and the resultant injection of non-borated l water from the VCT, is acceptable from the standpoint of long term core cooling considerations.

not lea Switchover to Prevent Potential Boron Precipitation - FSAR Chapter l

6.3.2.5 and FSAR Table 6.3-7 i

j Post-LOCA hot leg recirculation switchover time is determined for inclusion in' i emergency procedures to ensure no boron precipitation in the reactor vessel i following boiling in the core. This time is dependent on power level, and the

] RCS, RWST and accumulator water volumes and boron concentrations. A delay in j the injection of borated water to the RCS would have no impact on the power j level or volumes assumed for the RCS, RWST and accumulators, and will have l i negligible impact on the boron concentrations. Therefore, there will be i negligible impact on the post-LOCA hot leg switchover time for Beaver Valley

Unit 2.

CONCLUSION The proposed change in-the ESF' response times and table notations for i Containment Pressure - High 1, Low Pressurizer Pressure and Low Steamline Pressure in Technical Specification Table 3.3-5, Items 2.a. 3.a and 4.a to incorporate an increase of 15 seconds is acceptable. Evaluations of-the impact on the BV-2 safety analysi.s licensing basis, presented above,- '

demonstrate that the conclusions remain valid. A summary of the changes to 5

theBV-2DraftTechnicalSpecifidationsisprovidedonthenextpage, j .

REFERENCES

1. Beaver Valley Unit 2 (BV-2) FGAR e

1 2. WCAP-9220-P-A (Proprietary), WCAP-9221 (Non-Proprietary),

i Eicheldinger, C., " Westinghouse ECCS Evaluation Model - 1982. Version," '

Revision 1, 1981. ,

3. WCAP-10062 (Proprietary), Chiou, J.S., et al., "Models for PWR Reflood Calculations Using the.BART Code," March 1982.

l 4. WCAP-10054-P-A (Proprietary) . Lee, H. , et al., " Westinghouse Small - Break -

ECCS Evaluation Model Using the NOTRUMP Code," August 1985, I

i t. WCAP-8339 (Non-Proprietary), Bordelon, F.M., et al., " Westinghouse-ECCS Evaluation Model - Summary," June 1974.

!- 6. " Westinghouse Technf.:al Bulletin NSID-TB-86-08, " Post-LOCA Long-Term

! Cooling: Boron Reqdirescents," October 31, 1986.

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~ .. - ._ . ,, - _ . - ,- _.- . .-- .- . - . . - . - _ . _ _ _ . _ _

. -' ~ ATTACHMENT-C Page 6 BVP5-2 FSAR

- -l

b. The low head sis, and -

c. The high head SIS.

Only the high head system is modeled for the SI.B accident analysis. ,

The modeling of the SIS in LOFTRAN is described by Burnett (1972). The flow corresponds to that delivered by one SI pump delivering its full flow to the cold leg header. The L '

i water, which must be swept from the lines downstream of - the g, RWST prior to the delivery of concentrated boric acid to the is assumed to have no reactor coolant loops, concentration. boronf

< The calculation assumes the boric acid is mixed with and

! diluted by the water flowing in the RCS prior to entering p the reactor core. The concentration after mixing depends upon the relative flow rates in the RCS and in the SIS. The

, variation of mass flow rate in the RCS due to water density 4 changes is included in the calculation as is the. variation of flow rate in the SIS due to changes in the RCS pressure.

3 The SIS flow calculation includes the line losses in the system as well as the SI pump head curve. L.

The boric acid solution from the sis is assumed to be ,f i('

^

uniformly delivered to the active r'eactor coolant loops. ,

q i The boron in the loops is then delivered to the inlet plenum where the ceslant (and baron) from each loop is mixed and 3 delivered to the core. The stuck RCCA is conservatively assumed to be located in the core sector near the broken

, steam generator. Because the cold leg pressure is lowest in l , the broken loop due to larger loop flow and a larger pressure drop, more boron would actually be delivered to the g core sector where the power is being generated, enhancing e l the effect of the boric acid on the transient. No credit I was taken for this in the analysis. Furthermore, sensitivity studies have demonstrated that the transient is insensitive to boron worth or distribution. -

For the cases where offsite power is assumed, the sequence of events in the SIS is the following. After the generation 1 of the SI signal (appropriate delays for instrumentation, l logic, and signal transport included), the appropriate 1 l valves begin to operate and the SI pump starts. C In i' 12 seconds, the valves'are assumed to be in their final position and the pump is assumed to be at full speed. The volume containing the unborated water is swept into the core

! before the 2,000 ppm borated water reaches the core. This -

! delay, described herewith, is inherently included in the modeling.

E y ewg am e- y-w -+ - - -

rew -+-g ----g ,--er - er m ayg y +--,,m.,%m - -g-rw- w- ,,----w-gm yr-ym

~ unnfy ATTACHMENT C Page 7. ,

, TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

1. Manual

a. Safety Injection (ECCS) ,

Not Applicable j Feedwater Isolation Not Applicable Reactor Trip (SI) Not Applicable Containment Isolation-Phase "A" Not Applicable Containment Vent and Purge Isolation Not Applicable Auxiliary Feedwater Pumps. Not Applicable

] Service Water System Not Applicable i b. Containment Quench Spray Pumps Not Applicable containment Quench Spray Valves Not-Applicable Containment Isolation-Phase "B" Not Applicable

c. Containment Isolation-Phase "A" .

Not Applicable

d. Control Room Ventilation Isolation Not Applicable

^

2. Containment Pressure-High

a. Safety Injection (ECCS) 1 27.0*

1 i

b. Reactor Trip (from SI) 1 2.0

c. Feedwater Isolation < 7.0(1)

~

j t

d. Containment Isolation-Phase "A" 1 61.5(4)/115.5(5)

e. Auxiliary Feedwater Pumps 1 60.0

f. Service Water System 1 72.5(2)/181.5(3)

3. Pressurizer Pressure-Low p, g a. Safety Injection (ECCS) i 27.0*/na,#-

4

b. Reactor Trip (from SI) 1 2.0

c. Feedwater Isolation i 7.0(1)

d. Containment Isolation-Phase "A" 1 61.0(4)/115.0 (5)

BEAVER VALLEY - UNIT 2 3/4 3-29 l

O en

.} ,

ATTACIUIENT C Page 8. 90 pf TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS l 1

3. Pressurizer Pressure-Low (Continued)

e. Auxiliary Feedwater Pumps '

5 60.0

f. Service Water System 1 72.0(2)/181.0(3)

4. Steam Line Pressure-Low .

, 7 , M/, 7

[ a. Safety Injection (ECCS) 122. 0ff/12. %

b. Reactor Trip (from SI) 1 2.0

c. Feedwater Isolation 1 7.0(1)

d. Containment Isolation-Phase "A" 1 61.0(4)/115.0(5)

e. Auxiliary Feedwater Pumps 1 60.0

f. Service Water System 1 72.0(2)/181.0(3)

g. Steam Line Isolation 1 7.0

5. Containment Pressure--Nich-High

a. Containment Quench Spray 1 85.5(5) i

b. Containment Isolation-Phase "B" Not Applicable

c. Control Room Ventilation Isolation i 22.0(4)/77.0(5)

6. Steam Generator Water Level--Nich-High

! , s. Turbine Trip i 2.5 i

b. Feedwater Isolation 1 7.0(1)

7. Containment Pressure--Intermediate High-High

a. Steam Line Isolation 17.o

8. Steamline Pressure Rate--High Necative

a. Steamline Isolation i 7.0 1

d i

SEAVER VALLEY - UNIT 2 3/4 3-30 6

m_ _ . . . _ . -

. _ . - , , ., . . _ , , , - , _r, ._y_- ,., ,_ - _ ., - - , - . . _ . . - , -

7

~~

! ATTACHMENT'C' Page 9 FINAL D;IAFT TABLE 3.3-5 (Continued) J

\

TABLE NOTATION b i

N Diesel generator starting and sequence loading delays included. Response time limit includes opening of valves to establish SI path and attainment  %

of discharge pressure for centrifugal charging pumps and Low Head Safety Injection pumps. % y 7 j A ,

1. Diesel generator starting and sequence loadin'g delays not included. Offsite y power available. Response time limit includes opening 7 valves to establish s SI path and attainment of discharge pressure for centrifugal charging pumps. 1

1. 1. Diesel generator starting and sequence loading delays included. Response b y 3 time limit includes opening of valves to establish SI of discharge pressure for centrifugal charging pumps. ' path N and attainment (1) Feedwater system overall response time shall include verification of valve

~

stroke times applicable to the feedwater valves shown for penetrations 76,

. 77 and 78 shown in Table 3.6-1.

(2) Diesel generator starting and sequence loading delays included. Response

.. time limit includes attainment of discharge pressure for. service water pumps.

• (3) Diesel generator starting and sequence loading delays g included.

Response time limit only includes opening of valves to establish the flowpath to the diesel coolers.

(4) Diesel generator starting and sequence loading delays g included. Offsite I power available. Response ' time limit includes operation of valves / dampers. '

1 (5) Diesel generator starting, and sequence loading delays included. Response

time limit includes operation of valves / dampers.

fa.rcs r A'

l}

' feguenh*/ t'e- ~ sG e of e da y jg ,,,u n ,.. ,, a e pa n from 6 /e ver 4. g.4 e /%Jr ff w Jr- e a he.r y e a, & to s Vdf vs /#es e /ese ) p,p a./ is a /s. a(ee/.

h.2"%.seJ7

^

S' *

.reyae,D./ f ~ ~./e~ - a s c. 1 - - p , f u n, , a. t ) . ,.,

\

f,a A g/e ycy 4. -fd. A' A' 17 (A' " J 7 n her op e s , fle r VC7" vaher e /sse ) *t s /% e fuele d.

BEAVER VALLEY - UNIT 2 3/4 3-32 i .

i ATTACl!!!ENT C PACE in pr1NAL E

_3/4.3 INSTRUMENTATION 1

l BASES h 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM AND ENGINEERED SAFETY F ACTUATION SYSTEM IN5TRUMENTATION (Continued) ' j i The methodology to derive the trip setpoints is based upon combining all

! of the uncertainties in the channels. Inherent to the determination of the trip setpoints are the magnitudes of these channel uncertainties. Sensor and rack instrumentation utilized in these channels are expected to be capable of operating within tte allowances of these uncertainty magnitudes. Rack drift in excess met its allowance.of the Allowable Value exhibits the behavior that the rack ha will happen, an infrequent excessive drift is expected.8eing that there is in excess of the allowance that is more than occasional, may be indicative ofRa more serious problems and should warrant further investigation.

s Breakers, and Reactor Trip Bypass Breakers are provide bility of an Anticipated Transient Without Scram (ATWS) event by ensuring OPERABILITY of the diverse trip features (

Reference:

Generic Letter 85-09).

i The measurement of response time at the specified frequencies provides

assurance that the protective and ESF action function associated with each No credit was taken in the analyses for those channels wit indicated as not applicable.

m 4

d> - ywa y A fa . a n d M) Er #e.

or total channel test measurements provided that such tes total channel response time as defined.

i be demonstrated by either 1) in place, onsite or offsite test measuremen

2) utilizing replacement sensors with certified response times.

) { following The Engineered functions: Safety Feature Actuation System interlocks perform the P-4 Reactor tripped - Actuates turbine trip, closes main feedwater valves on T,,, below setpoint, prevents the opening of the main feedwater valves which were closed by a safety injection or high steam generator water can belevel resetsignal,or tripped. allows safety injection block so that components Reactor not tripped prevents manual block of safety injection.

P-11 1 Above the setpoint P-11 automatically reinstates safety injection (

q m actuation onM pressurizer pressure, automatically blocks steamline l isolation on high steam pressure rate, and enables safety injection 1 and steamline isolation on (Loop Stop Valve Open) with low steamline pressure.

1 i

BEAVER VALLEY - UNIT 2 8 3/4 3-2 4

0

.. ATTACPMENT_C. . - . _ . _ . ._ _ . . .. . .PAGE 11 O

{-

y f A.o_ .r s s r- A _ . __-- . . _ _ _ . _ _ . . . . .

. . . ESF response tians PMied in Table 3.3-5 eich irclude sequential . _ _ _ _ . _ . _

operation of the RHST arti VCT valves (Notes i and if) are based on values assumed in the non-IDCA safety analyses. 1hese analyses taka -- -

. credit for injection of borated water fra the RHyr. Injectim of _ . . . . . . .

borated water is ==M not to occur until the VCT chargirug puup suctim valves are ciceed following opening of the RHrr charging pump --

... _ _ . _ . . suction valves, men the sequential aparation of the RHET and VCP _ _.. __ ._ .

valves is not included in the resp:insa times (Nota *), the values specified are based cm the IDCA analyses. The IDCA analyses take - - - - - -

. credit for injection flow regardless of the source. Verification of the response times vMied in Table 3.3-5 will assure that the asstaptions used for the IDCA and Non-IDCA analyses with respect to - - - - - -

_ __ eparation of the VCT and RHEr valves are valid. _

1

.._8 . .. ,,

n - - - n , , , - . , - , . ,- ,-