ML20237J856

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Proposed Tech Specs Making Alternative Rod Insertion Design Mods to Comply w/10CFR50.62 ATWS Requirements
ML20237J856
Person / Time
Site: Grand Gulf 
Issue date: 08/13/1987
From:
SYSTEM ENERGY RESOURCES, INC.
To:
Shared Package
ML20237J853 List:
References
NUDOCS 8708180378
Download: ML20237J856 (15)
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Text

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4 REACTIVITY CONTROL SYSTEMS 3/4.1.5 STAND 8YLIQUIDCONTROLSYSTEM l

LIMITING' CONDITION FOR OPERATION l

- 3.1. 5 Two standby liquid control system subsystems shall oc OPERABLE.

l APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 5*.

ACTION:

a.

In OPERATIONAL CONDITION 1 or 2:

1.

With one system subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days or be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2.

With both standby liquid coritrol system subsystems inoperable, restore at least one subsystem to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b.

In OPERATIONAL' CONDITION 5*:

1.

With one system subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 30 days or insert all insertable control rods within the next hour.

2.

With both standby liquid control system subsystems inoperable, insert all insertable control rods within one hour.

SURVEILLANCE REQUIREMENTS

~

4.1.5.Each standby liquid control system subsystem shall be demonstrated OPERABLE:

a.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying that; 1.

The temperature of the sodium pentaborate solution is within the limits of Figure 3.1.5-1.

2.

The available volume of sodium pentaborate solution is greater than or equal to g allons.

l 3.

The heat tracing circuit is OPERABLE by determining the tempe'ature of the pump suction piping-t; be grcater--ther, er r

4q=1 to 70"T. is wWn the Ji W of Fiswe 315 1 m

"With any control rod withdrawn. Not applicable to control rods removed per Speci fica tion 3. 9.10.1 or 3. 9.10. 2.

kk DO P

1 RAND LULF-UNIT 1 3/4 1-18 bmendment No.

l REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b.

At least once per 31 days by; 1.

Starting both pumps and recirculating demineralized water to the test tank.

2.

Verifying the continuity of the explosive charge.

3.

Determining thtt the available weight of sodium pentaborate is I

greater than or equal to.550 ribs and the concentration of boron

)

in solution is within theflimitsofFigure3.1.5-1bychemical I

analysis.*

LS$00 4.

Verifying that each valve, manual, power operated or automatic, in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

c.

Demonstrating that, when tested pursuant to Specification 4.0.5, the minimum flow requirement of 41.2 gpm at a pressure of greater than or equal to le W psig is met.

Isoo d.

At least once per 18 months during shutdown by; 1.

Initiating one of the standby liquid control system subsystems, including an explosive valve, and verifying that a flow path from the pumps to the reactor pressure vessel is available by pumping demineralized water into the reactor vessel.

The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch which has been certified by having one of that batch successfully fired.

Both system subsystems shall be tested in 36 months. NC.

open6 d 2.

Demonstrating that he pump relief valve.setpciat is less

[

than or equal to.138frpsig and verifying that the relief valve q[,

does not actuate during recirculation to the test tank.

?$

EE-3.

    • Demonstrating that all heat traced piping between the storage "y

tank and the reactor vessel is unblocked by pumping from the storage tank to the test tank and then draining and flushing the piping with demineralized water.

4.

Demonstrating that the storage tank heater is OPERABLE by verifying the expected temperature rise for the sodium pentaborate solution in the storage tank after the heater is energized.

^This test shall also be performed anytime water or boron is added to the solution or when the solution temperature drops below the limit of Figure 3.1.5-1.

    • This test shall also be performed whenever both heat tracing circuits have been found to be inoperable and may be performed by any series of sequential, overlapping or total flow path steps such that the entire flow path is included.

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. INSTRUMENTATION

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3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.4.1 The anticipated transient without scram recirculation pump trip (ATWS-RPT) system instrumentation channels shown in Table 3.3.4.1-1 shall be OPERABLE with their trip setpoints set consistent with values shown in the Trip Setpoint column of Table 3.3.4.1-2.

APPLICABILITY:

OPERATIONAL CONDITION 1.

ACTION:

a.

With an ATWS-RPT system instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4.1-2, declare the channel inoperable until the channel is restored to OPERABLE status with the channel setpoint adjusted consistent with the Trip Setpoint value:

b.

With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels per Trip System requirement for one or both Trip System (s), restore the inoperable channel (s) to OPERABLE status within 14 days or be in at least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Insect SURVEILLANCE REQUIREMENTS 4.3.4.1.1 Each ATWS recirculation pump trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.4.1-1.

4.3.4.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.

it The inopeMle. c.bnnels need not be pixec) in he iripped c.ond%n u)here 4his. toouid caus.c +he. Trip Func.+ ion ft> oc.c.ac.

GRAND GULF-UNIT 1 3/4 3-37 Amendmed No, -

o I

4

' Insert To Page 3/4 3-37; c.

With;the number'of OPERABLE channels two or more. less than ' required.

I

- by the Minimum OPERABLE Channels per Trip. System requirement for one i

trip l system and:

~

1.

IfLthe inoperable channels consist of'one reactor ve'ssel water level channel and one reactor vessel pressure channel, place both inoperable channels.in the tripped condition

  • within one hour or declare the trip system inoperable.

2.

If the inoperable channels include two reactor vessel water level channels or two reactor vessel pressure channels, declare the trip systeni inoperable.

d.

With one trip system' inoperable, restore the inoperab!e trip system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

'e.

LWith both trip systems inoperable, restore at least one trip system to OPERABLE status within one hour or be in at least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

l l

L J16 MISC 87061502 --2 l

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TABLE 3.3.4.1-2 j

'l ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETP0INTS

[

TRIP ALLOWABLE TRIP iUNCTION SETPOINT VALUE

' 1.

Reactor Vessel Water Level -

-> -41.6 inches *

~> -43.8 inches Low Low, Level 2 b 1095psis

6. IlD3. ps,tg

{

2.

Reactor Vessel Pressure - High

,1 1125 psig

- 1 1140 pn g--

i

  • See Bases Figure B3/4 3-1.

i i

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GRAND GULF-UNIT 1 3/4 3-39 Amendm ent No, ~.---

^

I I

TABLE 3.6.4-1 (Continued)

CONTAINMENT AND ORYWELL !$0LATION VALVES SYSTEM AND PENETRATION JALVENUMBER

_ NUMBER Drw ell (6cntinued)

CRD to 1tectre.

833-F017A 326(0)

Pump A Seals Instrument Air PS3*F008 335(I)

Standby Liquid C41-F007 328(I)

Control

. Standby Liquid C41-F006 328/M($

Control e

Cont. Cooling P42-F115 329(!)

T Water Supply SteAy Lipid C.41 - FA1 7 32% (I)

Drywell Chilled P72-F147 332(I)

Codrol Water Supply Co ensate Flush E33-F204 333(I)

Stadby Licluid CMI-FAtB 32t(M Condensate Flush B33-F205 333(0)

\\

dentrol Conn.

Combustible Gas E61-F002A 339(0)

Control Combustible Gas E61-F002B 338(0)

Control.

Combustible Gas E61-F004A 340(0)

Control Combustible Gas E61-F0048 340(0)

Control Upper Containment G41-F265 342(0) 9001 Drain CRD to Rectre.

B33-F0138 346(I)

Pump B Seals CRD to Recire.

B33-F017B 346(0)

Pump B Seals Service Air PS2-F196 363(I)

Cont. Leak Rata M61-F021 438A(I)

Test Inst.

Cont. Leak Rate M61-F020 U A(0)

Sys.

1 BLIND FLANGES Cont. Leak Rata NA 40(!)(0)

Sys.

Cont. Leak Rate NA 82(I)(0)

Sys.

Containment NA 343(I)(0)

Leak Rata System GRAND GULF-UNIT 1 3/4 6-42 Amendment No. 21;.

_ _ _ _ _ _ _ _. _ - -. - - - - - - - - - - - - - ' - - - - ' - - - ' - ~ - -

7 m_

+

TABLE 3.6.4-1 (Continued)

CONTAINMENT-AND ORYWELL ISOLATION VALVES p

SYSTEM AND PENETRATION '

VALVE NUMBER-NUHBER Containment-(Continued).

RHR "B" Test Line E12-F350 67(0)(c)

T/C RHR "B" Test Line E12-F312 67(0)(C)

T/C-RHR "B" Test Line E12-F305, 67(0)(C)

T/C Refueling Water P11-F425 69(0)(c)

Transf. Pump Suction T/C-

' Refueling Water -

P11-F132 69(0)(C)

Transf. Pump Suction T/C

' Inst. Air to ADS P53-F043 70(0)

T/C-Post-Acc. Sample E12-F409 71B(I)(c)

Return and RHR "C" Relief' Valve Vent Hdr. to Suppr.

Pool.T/C Post Acc. Sample

.E12-F408 71B(0){c)

Return and.

RdR "C"

' Relief Valve Vent-Hdr. to Suppr.

Pool T/C Cont. Leak Rate M61-F010 82(I) j T/C RWCU To Feedwater G33-F055 83(0) 1 T/C Suppr. Pool P60-F011 85(0)

-Cleanup T/C Suppr. Poo?

P60-F034 85(0)

Cleanup I/C 1

RWCU Pump Suction G33-F002 87(0)

T/C i

RWCU Pump G33-F061 88(0)

Discharge T/C SSW T/C P41-F163A 89(0)(c)

SSW T/C P41-F163B 92(0)(c) b.

Drywell LPCI "A" T/C E12-F056A 313(0)

LPCI "B" T/C E12-F056B 314(0)

Instrument Air T/C P53-F493 335(0)

SLCS T/C' C41-5026%I5' 328(DRI)

Service Air T/C P52-F476 363(0)

RWCU T/C G33-F120 366(I)

Reactor Sample B33-F021 465(0)

T/C GRAND GULF-UNIT 1 3/4 6-45 Amendment No. 24 -

j

REACTIVITY C0!tTROL SYSTEMS e!-

BASES CONTROL R00 PROGRAM CONTROLS (Continued)

The RPCS provides automatic supervision to assure that out-of-sequence rods will. not be withdrawn or inserted. A rod is out of sequence if it does not meet the criteria of the Banked Position Withdrawal Sequence as described in the FSAR. The RPCS function is allowed to be bypassed in the Rod Action Control System (RACS) if necessary, for example, to insert an inoperable con--

trol rod, return an out-of-sequence control rod to the proper in-sequence l

position or move an in-sequence control red to another in-sequence position.

The requirement that a second qualified individual verify such bypassing and

. positioning of control rods ensures that the bases for RPCS limitations are not exceeded.

In addition, if THERMAL POWER is below the low power setpoint, additional restrictions are provided when bypassing control rods to ensure operation at all times within the basis of the control rod drop accident analysis.

The analysis of the rod drop accident is presedted in Section 15.4 of the FSAR and the techniques of the analysis are presented in a topical report, Reference 1, and two supplements, References 2 and 3.

The RPCS is also designed to automatically prevent fuel damage in the event of erroneous red withdrawal from locations of high power density during higher power operation.

A dual channel system is provided that, above the low power setpoint, restricts the withdrawal distances of all non peripheral control rods. This restriction is greatest at highest power levels.

3/4.1.5 STANDBY LIQUID CONTROL SYSTEM The standby liquid control system provides a backup capability for. bring-ing the reactor from full power to a cold, xenon-free shutdown, assuming that the withdrawn control rods' remain fixed in the rated power pattern. To meet this objective it is necessary to inject a quantity.of boron which produces a concentration of 660 ppm in the reactor core in approximately 90 to 120 min-453r utes. A minimum available quantity of*.468iP' gallons of sodium pentaborate g __ solution containing a minimum of M 1bs. of sodium pentaborate is required to meet a shutdown requirement of 3%. There is an additional allowance of 165 ppm in the reactor core to account for imperfect mixing and-tt.e fillinw of leda3e.

-ether - iping piem ccanected tu L reactor vc::cL The time requirement r

was selected to override the reactivity insertion rate due to cooldown follow-ing the xenon poison peak and the required pumping rate is 41.2 gpm. The min-imum storage volume of the solution is established to allow for the portion below the pump suction that cannot be inserted. The temperature requirement is necessary to ensure that the sodium pentaborate remains in solution.

1.

C. J. Paone, R. C. Stirn and J. A. Woolley, " Rod Drop Accident Analysis for Large BWR's," G. E. Topical Report NE00-10527, March 1972 2.

C. J. Paone, R. C. Stirn and R. M. Young, Supplemerit 1 to NE00-10927, July 1972 3.

J. M. Haun, C. J. Paone and R. C. Stirn, Addendum 2, " Exposed Cor s,"

Supplement 2 to NEDO-10527, January 1973 GRAND GULF-UNIT 1 B 3/4 1-4 Amendment No. 2sj l

b;,

REACTIVITY CONTROL SYSTEMS BASES' STANDBY LIOUID CONTROL SYSTEM (Continued)

With re'dundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to-continue ~for short periods of time with the system inoperable or for longer periods of time witn one of.the redundant components inoperable.

~

Surveillance requirements are established on a frequency that assures a high reliability of the system. Once the solution is established, boron con-centration will not vary unless more boron or water is added, thus a check on the temperature and volume once each 24 he.urs assures that the solution is available-for use, Replacement of th,e explosive charges in the valves at regular intervals will assure that these valves will not fail because of deterioration of the charges.

001scrt Num bec j 4

GRAND GULF-UNIT 1 B 3/4 1-4a Amendment No.23 _

I L_____1__ _______________ _______

i

Insert Number 1 To Page B 3/41-4a Compliance with the NRC ATWS Rule 10CFR50.62 has been i

I demonstrated by means of the equivalent control capacity concept using the plant specific minimum parameters.

This concept requires that each boiling water reactor must have a

standby liquid control system with a minimum flow capacity and boron t

content equivalent in control capacity to 86 gpm of 13%

weight j

sodium pentaborate solution (natural boron enrichment) used for i

the 251-inch diameter reactor vessel studied in NEDE-24222, Reference 4.

The described minimum system parameters (82.4 gpm, 13.6%

weight with natural boron enrichment) provides a

equivalent control capacity to the 10CFR50.62 requirement. Tho techniques of the analysis are presented in a licensing topical report NEDE-31096-P, Reference 5.

4.

" Assessment of BWR Mitigation of ATWS, Volume II",

NEDE-24222, December 1979.

5.

L.

B.

Claasen et.

al.,

" Anticipated Transients Without

Scram,

Response

to NRC ATWS Rule 10CFR50.62",

G.E.

Licensing Topical Report prepared for the BWR Owners' Group, NEDE-31096-P, December, 1985.

Page 1 of 1

INSTRUMENTATf0N BASES 1

ISOLATIONACTUATIONINSTRUMENTATION(continued) the A.C. power. supply is lost and is restored by startup of the emergency diesel generators.

In this event, a time of 10 seconds is assumed before the valve starts tc move.

In addition to the pipe break, t,he failure of the'D.C. operated valve is assumed; thus the signal delay (sensor response) is concurrent with the 10 second diesel startup. The safety analysis considers an allowsble inventory loss in each case which in turn detamines the valve speed in conjune-tion with the 10 second delay.

It follows that checking the valve speeds and the 10 second time for emergency power establishment will establish the re~sponse time for the isolation functions. However, to enhance overall system relia-bility and to monitor instrument channel response time trends, the isolation actuation instrumentation response time shall be measured and recorded as a part of the ISOLATION SYSTEM RESPONSE TIME.

Ope ~ ration with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equel t'o or greater than the' drift allowance assumed for each trip in the safety analyses.

3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling system actuatica instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the ability of the operator to control. This specification provides the OPERABILITY requirements, trip setpoints and response times that will ensure effectiveness of the systems to provide the design protection.

Negative barometric pressure fluctuations are accounted for in the trip setpoints and allowable values specified for drywell pressure-high. Although the instruments are listed by system, in some cases the same instrument may be used to send the actuation signal to more than one system at the same time.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or greater than the drift allowance assumed for each trip in.the safety analyses.

3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION pT a ic pa d ranfien wi eut cram TWS) cir ulat on p ip sys pr id s ns df 1 miti g th cons uene of eu ike oc urr nc of fai re to ser o rin an ntic ated ansi nt.

he spo eo th p1 t

t s ost lat d e ent alls ithin he e velop of tudy eve ts n hyk ra El et c omp ny pic Rep rt NE 10

,d d

rch 971 nd 00- 4 te D emb r 19, an Sect n 15 App ndix SA t

F R.

}

The end-of-cycle recirculation pump trip (EOC-RP'T) system 'is a part of the Reactor Protection System and is an essential safety supplement to the reactor trip. The purpose of the EOC-RPT is to recover the loss of themal margin which '

occurs at the end-of-cycle. The physical phenomenon involved is that the void reactivity feedback due to a pressurization transient can add positive reactivity to the reactor system at a faster rate than the control rods add negative scram reactivity.. Each EOC-RPT system trips both recirculation pumps, reducing coolant '

flow in order to reduce the void collapse in the core during two of the most limiting pressurization events. The two events for which the EOC-RPT protective GRAND GULF-UNIT 1 B 3/4 3-2 Amendment No.

19,__

N

4.

L l

I Insert Number 2 To Page B 3/4 3-2 The anticipated transient without scram recirculation pump trip (ATWS-RPT) system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient. -The response of the plant to this postulated event has been evaluated in General 1

Electric company report NEDC-32408 dated March, 1987. The results of the analysis show that the Grand Gulf ATWS-RPT design provides adequate protection for these events in which the normal scram paths fail.

The ATWS-RPT provides fully redundant trip of the recirculation pump motors so that the pumps coast down to zero speed. This trip function reduces core flow creating steam voids in the core, thereby decreasing power generation and limitino any power or pressure excursions.

i J16 MISC 87072402 - 1 f

. to AECM-87/0152 l

Brief Description of the Grand Gulf Nuclear Station Alternate Pod Insertion (ARI) System

]

1 1.

10CFR50.62 requires that each boiling water reactor (BWR) implement.

certain design features to mitigate the consequences of an anticipated transientwithoutscram(ATWS) event. The required features are 1) aa alternate rod insertion (ARI) system, 2) a standby liquid control system i

(SLCS) with a minimum flow capacity and boron content equivalent in control capacity to 86 gpm and 13% weight sodium pentaborate solution, and

3) an automatic recirculation pump trip (RPT) system.

In AECM-85/0322 dated October 14, 1985, Grand Gulf Nuclear Station (GGNS) committed to

'instcll an ARI system, modify the SLCS system to accommodate simultaneous operation of both pumps to achieve the required equivalent control capacity, and modify the existing RPT system for energized-to-trip logic.

Subsequently, General Electric issued Licensing Topical Report NEDE~31096-P for the BWR Owners' Group.

i 2.

In MAEC-87/0039 dated February 11, 1987, the NRC stated that the information-in AECM-85/0322 was not sufficient to demonstrate compliance with the requirements of 10CFR50.62. The NRC requested addition 31 GGNS ATWS design information and that GGNS specifi.a' endorse the designs described in NEDE-31096-P and the associated NRt. ;,afety evaluation. GGNS provided the requested design information in AECK 87/0055 dated April 3, 1987 and committed to the designs described in NEDE-31096-P and the applicable conditions of the NRC safety evaluation.

3.

The ARI system to be implemented at GGNS consists of three parallel vent paths from the scraw pilot air header.

Each vent path consists of two solenoid valves in series for a total of six vent valves.

(See Figure 1).

These solenoid valves are normally de-energized and are energized from redundant ARI scram initiation trip systems to open and deprescurize the scram pilot air header.

Each ARI scram' initiation trip syster is tripped on conditions indicatio of an ATWS event (RPV pressure high cr RPV water level low).

The trip logic is two-out-of-two for pressure or two-out-of-two for level.

(See Figure 1 of Attachment 3). The trip function may also be initiated manually.

4.

The ARI system uses the same setpoints and trip channels (transmitters and trip units) as the RPT system. Therefore, the ARI trip function and the RPT trip function will Le initiated simultaneously. The instrumentation setpoints for the RPV pressure and water level trip channels are established j

such that the normal scram paths for these variables would already be initiated.

5.

The transmitters, trip units and output relays are shared in common by the ARI and RPT systems. Therefore, the RPT monthly functional test required by current Surveillance Requirement 4.3.4.1.1 will also be sufficient to adequately demonstrate the operability of the ARI/RPT shared components.

Additionally, through plant administrative procedures the ARI valves will be functionally testeo every 18 months to demonstrate ARI valve actuation.

J16 MISC 87061502 - 1

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