ML20086B706

Proposed Tech Specs Re Increasing Required RPV Boron Concentration & Modifying Surveillance Frequency for SLC Pump Operability Testing

ML20086B706
Person / Time
Site:	Cooper
Issue date:	06/28/1995
From:	NEBRASKA PUBLIC POWER DISTRICT
To:
Shared Package
ML20086B704	List: ML20086B702 ML20086B706
References
NUDOCS 9507060063
Download: ML20086B706 (26)

Text

.o RADIOLOGICAL TECHNICAL SPECIFICATTONS

, TABLE OF CONTENTS ,

Page No.

-1.0 DEFINITIONS 1-5 LIMITING SAFETY SAFETY LIMITS SYSTEM SETTINGS 1.1 FUEL CLADDING INTEGRITY 2.1 6 - 22 [

1.2 REACTOR COOLANT SYSTEM INTEGRITY 2.2 23 - 26 SURVEILLANCE LIMITING CONDITIONS FOR OPERATION REQUIREMENTS 3.1 REACTOR PROTECTION SYSTEM 4.1 27 - 46 3.2 PROTECTIVE INSTRUMENTATION 4.2 47 - 92 A. Primary Containment Isolation Functions 47 B. Core and Containment Cooling Systems Initiation 47 and Control (CS, LPCI, HPCI, RCIC, ADS)

C. Control Rod Block Actuation 47 D. Radiation Monitoring Systems - Isolation and 48 Initiation Functions

1. Steam Jet Air Ejector Off-Gas System 48

2. Reactor Building Isolation and Standby Gas 48 Treatment Initiation

3. Liquid Radwaste Discharge Isolation 48

4. Main Control Room Ventilation 48

5. Mechanical Vacuum dump Isolation 48 E. Drywell Leak Detection 49 F. Primary Containment Surveillance Information 49 Readouts G. Recirculation Pump Trip 49 H. Post-Accident Monitoring 49

1. Alternate Shutdown Capability 49 3.3 REACTIVITY CONTROL 4.3 93 - 106 A. Reactivity Limitations A 93 B. Control Rods B 94 C. Scram Insertion Times C 97 D. Reactivity Anomalies D 98 E. Restrictions E 98 F. Recirculation Pumps F 98 G. Scram Discharge Volume G 98a 3.4 STANDBY LIQUID CONTROL SYSTEM 4.4 107 - 113 A.

p.

Normal Operation

[ System]

A 107 B. Operation with Inoperable Components B 108 C. Sodium Pentaborate Solution C -M&- 10 7 i

9507060063 950628 11/o7fgg PDR ADOCK05000gB P

j SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS i 1.1- FUEL CIADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY Apolicability Anolicability .

The ISafety Limits established to Th Limiting Safety System Settings) '

preserve the fuel cladding integrity  ; ' apply to trip settings of the  !

apply to. those variables which instruments. and devices which are l monitor the fuel thermal behavior. > provided to prevent the fuel _ i J\ cladding integrity  !

gb from bein5 exceeded.Qafety Limits)  ;

Obiective l

Obiective The objective of the [ Safety Limit) s m is to establish limits below which The objective of theEtimiting Safetv )

the integrity of the fuel cladding r[ System Settings} is to define the is preserved. l level of the process variables at j j which automaticfprotective action)1s ,

Ag_ti2D

[ initiated cladding to prevent the fuel _

If a [ Safety Limit [is exceeded, the integrity "E******d(Safety Limits]

l reactor shall be in at least N M i (shutdownhithin2 hours. \ Specifications Specifications '

A. Reactor Pressure >800 nsia A. Trin Settings  !

p_nd Core Flow >10% of Rated The existence of afminimum )

The 4+it-NW Mm wM settingsJ shall be as '

feritical power ratiof(MCPR) specified below:

less than 1.06 for two recirculation loop operation 1. Neutron Flux Trin Settines (1 07 for single-loop operation) shall constitute a. APRM Flux Scram Trin Setting  !

violation of the fu 1 (Run Mode) cladding integrity (safety]

Core lT. ) (1) Flow Referenced Scram Trin .

l l

B. Thermal Power Lim t Setting ggM ,

_(Reactor Pressure $800 nsia When the Reactor Mode Selector and/or Core Flow (10%

is in the RUN position, the  !

When the reactor pressure is APRM flow referenced flux  ;

(800 psia or core flow is scram trip setting shall be-  !

less than 10% of rated, the core thermal power shall not Ss;0.58 W + 62% - .58 AW I l

exceed 25% of [ rated) th;c. .1 where:  !

(P -

d S - Setting in nercent of  !

C.

Power Transient 7 g 7

@ .,(rated n 1 power]

o ensure that the y ( " )O Qimit) established Specification 1.1. A and 1.1.B in W - Two-loop recirculation  !

flow rate in percent of l is not exceeded, each rated (rated loop '

required scram shall be ini- recirculation flow rate '

is that recirculation flow tiated by its signal. Theexpected (Safety Limit scram [ rate which provides 100%

shall be assumed to be exceeded when scram is coreflowat'100%powerg l  !

accomplished by a means other AU - Difference between-than the expected scram two-loop and single-loop signal. effective drive flow at the same core flow.

11/29/91

? .

[" 1.1 Bases: (Cont'd) All Co-(>S C. Power Transient

~

Plant safety analyses have shown that the scrams caused by exceedinz any safet setting will assure that theQafety Limifof Specification 1.lA or 1.lB will not be exceeded. Scram times are enecked periodically to assure the insertion times are adequate. The thermal power transient resulting when a scram is accomplished other than by the expected scram signal (e.g. , scram from neutron flux following closure of the main turbine stop valves) does not ne_cessarily i.

cause fuel damage. However, for this specification aCSafety Limit;Tviolation -

will be assumed when a scram is only accomplished by means of a backup, feature J) of the plant design. The concept of not approaching a(Safety Limith rovided scram signals gperable)is supaorted by the extensive plant safety analysis.

CheMr ShkoD The computer provided with Cooperthas a sequence annunciation program which will indicate the sequence in which events such as scram, APRM trip initiation, pressure scram initiation, etc. occur. This progran also indicates when the scram setpoint is cleared. This will provide information on how long a scram condition exists and thus provide some measure of the energy added during a transient. Thus, computer information@--ljwill be available for analyzing scrams; however, if the computer informationqshould not be available for any l

A

( -

simit ramhasanalysis, Specification 1.1.C will be relied on to determine if aQ been violated. j m

normalh 9 D. Reactor Water Level (Shutdown Condition)

During periods when the reactor is onsideration must also be given to water level requirements due to the etiect of decay heat. If reactor water

) level should drop below the top of the active fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation. The core can be cooled sufficiently should the water level be reduced to two-thirds the core height. Establishment of tHe (safety limitJat 18 inches above'the top of the fuel provides adequate marg n. _.

a.lt References for 1.1 Bases [ fps W

1. " General Electric Standard Application for Reactor Fuel,"

NEDE-240ll-P-A-(latest approved revision).

2. " Cooper Nuclear Station Single-Loop Operation," NEDO-24258, May, 1980, 04/12/90

e a\\

cgo NOTES FOR TABLE 3.1.1

'l. There shall be twjoperable]

o or tripped itrip systems) for each function. _

If the [

minimum number of(operable instrument channels] for a ttrip system 3cannot be met, the affected (trip systemyshall be placed in the safe (tripped) condition, or the appropriate actions listed below shall be taken. a f

A.. Initiate insertion of(operable} rods and complete insertion of all} operable]/ rods within four hours. ~

E. Reduce power to less than 30% of hate _

g]

Reduce power level to IRM rangeQeactor) ana placekdeWw. itch in theIStartup position

~

~g E

. (t C.

within"/ hours and depressurize to less than 1000 psig. ~

l

2. Permissible to bypass, with control rod block, for Reactor Protectionsystem reset i in[iefuegd(shutdown)positionsofthekeactor@ ode / witch.

3. This note deleted. [ Selector)

4. Permissible to bypass when turbine first stage pressure is less than 30% of full load. {gegefo]

5. IRM# s are bypassed when APRM # s are onscale and the keactorkode witch is in the run position.

6. The design permits closure of any two lines without a full scram being initiated.

7. When tha reactor is subcritical, fuel is in the vessel, and the reactor water temperature is less than 212*F, only the following trip functions need to be

-(ojierablep hw eleckor'1

~

/

Veae.1or -

. Mod Switchin(shutdowj. g gy. Manual scram.

~

u C r. IRM high fluxf b20/125indicatedscale. Y-p.#. APRM (15%) high flux scram.

8. Not required to be (operable} whenfrimary containment incegrity)is not required,

~ -

f o be OPERABLO

9. Not requiredphile w performing low power physics tests at atmospheric pressure during or after refueling at power levels not to excee'd 5 MW(t).

10. Not required to bet 6perableJEen the reactor pressure vessel head is not bolted to the vessel.

I h, ke_ duct WC M O a,nd C'kO5Q , od t\ Q W)

Isot Mon W.Wes M44n eight hcors.

3/2/93

.m: -

1.s s

4 w >

COOPER NUCLEAR STATION TABLE 4.2.C SURVEILIANCE REQUIREMENTS FOR ROD WITHDRAWAL BLOCK INSTRUMENTATION Functional Function Test Freo. Calibration Freo. Instrument Check-APRM Upscale (Flow Bias) (1) (3) Once/3 Months Once/ Day APRM Upscale (Startup Mode) (1) (3) Once/3 Months ,

.Once/ Day APRM Downscale (1) (3) Once/3 Months Once/ Day-APRM Inoperative (1) (3) N.A. Once/ Day RBM Upscale'(Power Referenced) (1) (3)- Once/6 Months Once/ Day RBM Fower Range (3) Once/6 Months N.A.

RBM Downscale. (1) (3) Once/6 Months Once/ Day RBM Inoperative (1) (3) N.A. 'Once/ Day-IRM Upscale (1) (2) . (3) Once/3 Months .Once/ Day.

IRM Downscale (1) (2) (3 Once/3 Months Oncd/ Day-IRM Detector Not Full In (2 (Once oper- Once(oper.Cyc1 (10) Once/ Day ating c cle)

IRM Inoperative (1) (2) (3) N.A. N. A T SRM Upscale (1) -(2) (3) a0 Once/3 Months Once/ Day 0 SRM Downscale (1) (2) (3 C"@ Once/3 Months Once/ Day 7 SRM Detector Not Full In (2) (Once oper- Onceker. Cycle (10)- N.A.

! (atingcycle)

SRM Inoperative (1) - (2) (3) N.A. N.A.

Flow Bias Comparator- (1) (8) OnceMper. Cycle _l N.A.

Flow Bias Upscale (1) (8) Once/3 Months N.A.

Rod Block Logic (9) N.A. .N.A.

20C5 M ;;; (l' ^nc ;/3 M ac.t'.; "..".

SDV High Water Level Quarterly- Onceglper. Cycle) N.A.

S

- god M . CAP 3hara+2^$

. i.

i 9

D ~

R .

~w---+s m,,-v-,-en-+r vs,-r-s--,m-~~-,r~+, s w e e- ,-,,-me+>- -~w~-s + ~ --w-r--w+ m-s v - v,~, ev-w ~r- * ~ ~ = - r - -' a '~~v. -

ev e~ s~~w---N------~-+------'------

. LIMITING CONDITIONS FOR OPERATION SURVEILIANCE REOUIREMENTS 3.4 STANDBY LIOUID CONTROL SYSTEM 4.4 STANDBY LIOUID CONTROL SYSTEM Aeolicability: Avelicability:

Applies to the operating status of Applies to the surveillance the Standby Liquid Control System. requirements of the Standby Liquid Control System.  !

(SLC)] (SLC)) all caps, l 1

Obiective: Obiective OPERABluTY)

To assure the -' ' ^ ili ty of a To verify the [ operability] of the y system with the capability to -Ctr e; '_iq;id C:ntr:1 System.

caps M[ shutdown (he reactor and maintain J i

Q (shutdown) condition without the use of control rods.

Specification: Specification: -

[ Ope.rakon ) gbOperation]

A. Normal System !.. ..i1 '.111; 7 .

A. Normal System " ail iili j 1, During periods when fuel is in the reactor and prior to startup from a j p4f3 h c ra m M of de e " '-

, , _,, s _ .- '- System shall be shown Cold Condition, the Standby Liquid q b[theperfbanceofthefollowing Control System shall be operable,

• • 8 except as specified in 3.4.B below.

This system need not be operable ch 3 months)

1. M - -" ed when the reactor is in the Cold '

be tested for Condition and all control rods are g g l

)

fully inserted and Specification 3.3.A is met, demineralized water to the test tangand veMGpe cach PwmP develops a 4:t o , rate. E 3 8.7. gym a t o, di scha cy, pressure E GOO P5 3

2. Atleastonceduringeachoperating}

all caps

a. Check that the settings of the l s stem relief valves are l

< 1450 ( P( 1680 psig and the valves w eset at P 2 1300 psig.

b. Manually initiate the system, except explosive valves, and pump boron g g solution from the f .. 1 ) Liquid Gm al- Storage Tank through the recirculation path. "ir i-- pq l Mid, ,a r n f j ,[_

a / '"

After pumping boron N/e.r~fy each pump d eve _ lops solution the system will be flushed

{ low rake >_ 3 8,'l gm ab o with demineralized water.

disc ho.rge pressure h t3OO esT3.

L c. Manually initiate one of the C ney \

...O d I '[st.c q"' ' *f System Pumps an j

> m ove to R t. 108

-107- 3/11/92

1 b

4

.. . LIMITING CONDITIONS FOR OPERATION SURVETLTANCE REOUTRFRFETS i 3 . 4' 4.4.A.2 Cont'd.)

pump demineralized water into the reactor vessel from the test tank.

These tests check the actuation of the explosive charge of the tested Sd h 'W, proper operation of

( valves, and pumploperability/, The

~

the r

replacement charges to be installed will be selected 'from the same '

manufactured batch as a previously.

L$CT0fR tested charge.

T)0HER s d. Both subsystems, including ~ both l

' explosive valves, shall be tested in d the course of g operating cycle [

B. Oneration with Inocerable C>N B. Surveillance with Inocerable Components:  !

Comoonents:

1. From and after the date that one 1. When a subsystem is found to be subsystem is made or found ' to be 9 in Perable, thelparablej subsystem' '

{ inoperable, Specification ' 3.4'. A.1 shall be verified to be (operable )

\ shall be considered fulfilled and continuedi Coperatio7 permitted immediately and daily thereafter '

provided that thedsperablifsubsystem until the inoperable _ subsystem is remainsCoperablefsnd the inoperablej returnedtoanoperabgcondition. ]

l subsystem is returned to an@erable')

- condition within seven days.

C. Sodium Pentaborate Solution r s C. Sodium Pentaborate Solution i g{SLC) '

At all times when the ., .._, me- The .following tests shall be  !

-Cea rel System is required to be performed to verify the availability J g . gperable) the following conditions ofthe/iquid[ontrolfolution:

g shall be met:

1. The net volume versus concentration 1. Volume: Check and record at least of the [iquid[gontrol /olution in once per day, li,__8 d - centrol [ank shall be maintained as required in 4 9 ,g i Figure 3.4.1.

I  !

2. The temperature of the liquid 2. Temperature: Check and record at l contro1 solution shall be maintained least once per day. i above the curve shown in Figure 3.4.2. 3. Concentration: Check and record at least once per month. Also check concentration anytime water or boron m I ev-R. h j P ^ y 1 0 ')

-108- 3/11/92 i

. ri! '

y ' '.-,14

+=

• i

'I I

~

.,,,,,, ' LIMITING CONDITIONS FOR OPERATION f SURVEILLANCE REQUIREMENTS '

, , 3.4' 4.4.Cid 's ..;'d.) (b' L

added to' the solution or solution i i

temperature is below the temperature i required in Figure 3.4.Eg)

'D. {

If specification 3.4.A through C t cannot be met, the reactor shall be i placed in a' Cold Shutdown Condition  :

8 with all operable control rods fully ;

inserted within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. -

j l

i i

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i

%s t

t l

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'l

-i

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j 1

-109-

3.4 BASES iATED POWERl L ,

STANDBY LIOUID CONTROL SYSTEM (SLC S LC.

% A. The Standby Liquid Control System consists of two, distinct subsystems, each SLC Stomje containing one positive displacement pump and independent suction from the

.iquid c;ntrol tank, and discharge to a common inj ection header through (explosive) parallel *:quibb valves. The purpose of the Standby Liquid Ccatrol" System is i

to provide the capability of bringing the reactor from full peu r"to a cold,

((b3gh xenon-fret [ shutdown condition) assuming that none of the withdrawn control rods can be inserted. To meet this objective, the system is designed to inject a l quantity of boron that produces a concentration of -60Vppm of boron in the reactor .eese in less than 125 minutes. The -606# ppm concentration in thet j

~

reacto seregis required to bring the reactor from full p:::r'to a 3.0 percent Ak subcritical condition, considering the hot to cold reactivity difference,

%' VO xenon poisoning, etc. The time requirement for inserting the beron solution was selected to override the rate of reactivity insertion caused by cooldown of the reactor following the xenon poison peak.

S' LC The conditions under which the "t:ndby Li,- 2 C;ntrel System must provide shutdown capability are identifie 1: the "lant "uclear S:fety Operational l Analytic (App:hdi. C) . If no more than oneCoberablercontrol rod is withdrawn, q Standby Liquid Centrol System is not required. Thus, the basic reactivity SLC requirement for the core is the primary determinant of when the ';tm.dh,7 Liquid Centrol System is required. - -

ggfg g The minimum limitation on the relief valve setting is intended to prevent the recycling of liquid control solution via the lifting of a relief valve at too low a pressure. The upper limit on the relief valve setting provides system l protection from overpressure.

3. Only one of the two Standby Liquid Control subsystems is needed for o erating l the system. One inoperable subsystem does not immediately threaten shutdown capability, and reactor operation can continue while the inoperabic subsystem is being repaired 4 A:: uran:: .h :t th: re= ining r"Myr t- "ill perferr its intended f_tien ond thot the long-term-average-availebi44ty-of-the-system-is net-reduced-is--obtained-for a en aut of tre syster by an-ellowable-equipment eut a ee mica t= 4 af- ene-third-of-the-normal-survel14ance-frequeneyr -This-e d ad determinar e quipment-4ut-of-senic a time-of-cen-dayst--Addit 4enal-oonservatism-is-4ntroduced-by-reducirrg-the-allowable-out-efnervice ti- to

.saven--days-I (C. Level indication and alarm indicate whether the solution volume has changed, which might indicate a percibl# solution concentration change. The test interval has been established in consideration of these factors. Temperature and liquid level alarms for the system are annunciated in the control rc;m.

The solution is kept at least 10*F above the saturation temperature to guard against boron precipitation. The margin is included in Figure 3.4.2.

N moucu % pam \\\ q

(%. seven ekd&me h had en k avabSOb' an OPERAB E s sus +em co hG of m the k LQ s s+em p+mk au% Lwr &4U ^3e# Bass Aa%H~

110'- [

[3/)$f Scl&L6 band .

ecc v wUA 'M of /hoOm&M&/(dRD)3/11/92 sfe m %q conds wrn % p ad.j -

?.

.56c.T I O h) FRb%

/ M E /10 m ovGD-3.4 BASES,(cont'd.) NM _

The volume versus concentration requirement of the solut!.on is such that, should evaporation occur from any point within the curve, a low level alarm j

will annunciate before the temperature versus concentrat:lon requirements are exceeded.

The quantity of stored boron includes an additional margin (25 percent) beyond the amount needed to shutdown the reactor to allow for possible C.om0[4g PARA & RAP 6 imperfect mixing of the chemical solution in the reactor water. M d al32.1

, A minimum quantity of fe40 gallons of. solution having a 16.0 percent sodium pentaborate concentration, or the equivalent as shown in Figure 3.4.1, is (requiredtomeetthisshutdownrequirement.

The NRC's final rule on Anticipated Transients Without Scram (ATWS),

10CFR50.62, req.sf res that the-seendby 1iquid Ccatsc185y stem (CLC4 be modified to provide a minimum flow capacity and boron content eqaivalent in control capacity to 86 gpm of 13 weight percent sodium pentaborate solution for a 251 inch I.D. vessel. For Cooper Nuclear Station, with_a 218 inch I.D.

vessel, the equivalent minimum flow rate is 66 gpm e #l3 weight percent sodium pentaborate. This equivalence is met with both SLC/ ptmps supplying their minimum flow rate of 38.2 gpm with a solution concentra tion of at least '

11.5 weight percent of sodium pentaborate. Because ATUS is a very low probability event and is considered to be beyond the dusign basis for CNS, the surveillance and limiting condition for operation requgrements need not b_e___

more stringent than the original SLC/* design b sis requirements. The SLCf O changes made as a result of the ATWS rule do not inval-ldate the original system design basis.

4.4 BASES

[]

f 1, fesT.g once. each STANDBY L1 QUID CONTROL SYSTEM N *^b j Experience with pum)[operabilitylindicates that -t :d!;-east , in combination g with the tests during eachloperating cycle /, is sufficient t o maintain pump g' caps Performance._ The only practical time to fully test the 444 tid contro1* system . - - _ ,

is during a~(refueling outaggf. Various components of the system are individually tested periodically, thus making unnecessary more frequent testing of the entire system.

SAR The bases for the surveillance requirements are given in Mection{'I III . ^ . h of-- s ,

=th; Tinal-6ef+tt-Analycic Sporti and the details of the vyious tests are E-9 #5 discussed in suds'ection III.^.5 F The solution temperature and volume are

~

checked at a frequency to assure a high reliability of operation of the system should it ever be required.

s'

-111-07/05/88

17 /*

il>laligigigilsjil: gilijijil lijajajijijijijijajil1ltlsjijll

_ Maximum Concentration 16*/o _

_ Overflow vol. ~

_ 45pgol. _

15 % -

e 2: .

.e _

5 g -

~

N / -

14 % -

i$

~

~ .s 99C $, -

E M Reg 10n of Allowable -

vs IS % -

~

h Vol.- Conc.

.5 -

g, -

12 % ~

Minimum Concentrollon .

11 %

2000 'I ' I ' I ' I 2500

' I ' I ' I ' I ' I 3000

' I ' I ' I ' I ' I ' I ' I ' ' ' l ' ' ' I ' I ' I ' ' ' ' ' 3. ' I ' ' ' I ' I ' l _

3500 4000 4500 5000 Net Tonk Volume (gol.)

NE8RA5KA PUBLIC POWER DISTRICT Co0PER NUCLEAR STATION Sodium Pentaborate Solution Volume-Concentration Requirements FIGURE 3.4.1 112 07/05/88

i f ,

' e(

.. l I

'T BASES: '

(--

3.6.H and 4.6.H .

1 Snubbers- f i

Snubbers are- designed to prevent unre' strained pipe motion under dynamic loads as might' '

sgg - cecur during an earthquake or severe transient, while allowing normal. thermal motion  ;

p __ uring (startup)an4fshutdowrp. The consequence of an inoperable snubber is an increase in -

' ,W the prob'abl % f structural damage to piping as a result of a seismic or other event ...

initiating dynamic loads. It is therefore required that all snubbers required to protect  ;

the primary coolant system or an other safety system or. component be glh during l

' h*$' 4 ca-PS _

Bscause the snubber protection is required only during relatively low probability events a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed for repairs or replacement. Since plant startu s ou d not- commence prohibits withthknowingly defective safety related equipment Secificaton3.6.H[

inoperable snubbers.

All safety related snubbers are visually inspected for overall integrity and operabilit}y. l The inspection frequency is based upon maintaining a constant level of snubber protection.  !

Thus the required inspection interval varies inversely with the observed snubber failures.  !

The number of inoperable snubbers found during a required inspection determines the time  !

interval for the next required inspection. Inspections performed before' that interval has elapsed may be used as a new reference point to determine the next inspection.

However, the results of such early inspections performed before the original required time >

i interva).: has elap' sed (nominal time less 25%) may not be used to lengthen the required  !

( inspection interval. Any inspection whose results require a shorter inspection interval I will override the previous schedule.

When the cause of the rejection of a snubber is clearly established and remedied for that j snubber and for any other snubbers that may be generically susceptible, and verified by l  ;

functional testing, that snubber may be exempted from being counted as inoperable.  !

Generically susceptible snubbers are those which are of a specific make or model and have the same design features directly related to rej ection of the snubber by visual  ;

inspection, or are similarly located or exposed to the same environmental conditions, such i as temperature, radiation and vibration. i

.l l

-152- 10/31/91

i

.. i

'e" Attachm:nt 1 to NLS950115 Page 9 of 9 l-

?

t APPENDIX B REVISED TECHNICAL SPECIFICATIONS PAGES i

a

RADIOLOGICAL TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Pare No.

1.0 DEFINITIONS- 1-5 LIMITING SAFETY SAFETY LIMITS SYSTEM SETTINGS p 1.1 FUEL CLADDING INTEGRITY 2.1 6 - 22 f 1.2 REACTOR COOLANT SYSTEM INTEGRITY 2.2 23 - 26 SURVEILLANCE LIMITING CONDITIONS FOR OPERATION REOUIREMENTS 3.1 REACTOR PROTECTION SYSTEM 4.1 27 - 46 3.2 PROTECTIVE INSTRUMENTATION 4.2 47 - 92 A. Primary Containment Isolation Functions 47 B. Core and Containment Cooling Systems Initiation 47 and Control'(CS, LPCI, HPCI, RCIC, ADS) '

C. Control Rod Block Actuation 47 D. Radiation Monitoring Systems - Isolation and 48 Initiation Functions

1. Steam Jet Air Ejector Off-Gas System 48

2. Reactor Building Isolation and Standby Gas 48 Treatment Initiation

3. Liquid Radwaste Discharge Isolation 48

4. Main Control Room Ventilation 48 )

5. Mechanical Vacuum Pump Isolation 48 E. Drywell Leak Detection 49 F. Primary Containment Surveillance Information 49 Readouts G. Recirculation Pump Trip 49 H. Post-Accident Monitoring 49 ,

I. Alternate Shutdown Capability 49 3.3 REACTIVITY CONTROL 4.3 93 - 106

a. Reactivity Limitations A 93 B. Control Rods B 94 C. Scram Insertion Times C 97 I D. Reactivity Anomalies D 98 E. Restrictions E 98 F. Recirculation Pumps F 98 G. Scram Discharge Volume G 98a 3.4 STANDBY LIQUID CONTROL SYSTEM 4.4 107 - 113 A. Normal System Operation A 107 l B. Operation with Inoperable Components B 108 C. Sodium Pentaborate Solution C 109 l

.i.

I

. i SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS 1.1 FUEL CLADDING INTEGRITY 2.1 FUEL CIADDINC INTEGRITY Anolicability Applicability l The SAFETY LIMITS established to The LIMITING SAFETY SYSTEM SETTINGS l preserve the fuel cladding integrity apply to trip settings of the apply to those variables which instruments and devices which are monitor the fuel thermal behavior. provided to prevent the fuel l cladding integrity SAFETY LIMITS l Obiective from being exceeded.

l The objectf ve of the SAFETY LIMITS Obiective is to establish limits below which the integrity of the fuel cladding The objective of the LIMITING SAFETY is preserved. SYSTEM SETTINGS is to define the level of the process variables at Action which automatic PROTECTIVE ACTION is l If a SAFETY LIMIT is exceeded, the initiated to prevent the fuel reactor shall be in at least HOT 1 adding integrity SAFETY LIMITS l SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, fr m being exceeded.

Specifications "*# * "*

A. Reactor Pressure >800 nsia A. Trio Settines and Core Flow >10% of Rated The existence of a MINIMUM

• " ^

l SETTINGS shall be as specified l CRITICAL POWER RATIO (MCPR) below:

less than 1.06 for two recirculation loop operation

1. Neutron Flux Trio Settines (1.07 for single-loop operation) shall constituts the a. APRM Flux Scram Trio Settine violation of fuel cladding integri ty SAFETY l LIMIT. (1) Flow Referenced Scram Trio B. Core Thermal Power Limit y

(Reactor Pressure <800 nsia When the Reactor Mode Selector and/or Core Flow <10%) Switch is in the RUN position, l l When the reactor pressure is the APRM flow referenced flux l s ram trip setting shall be:

<800 psia or core flow is less than 10% of rated, the Ss0.58 W + 62% - .58 AW j core thermal power shall not '

l exceed 25% of RATED POWER. i C. Power Transient S - Setting in percent of RATED POWER (2381 MWt). l To ensure that the SAFETY LIMIT established W - Two-loop recirculation in Specification 1.1. A and 1.1.B flow rate in percent of '

exceeded, rated (rated loop is not each required scram shall be rehulde fb me is that recirculation initiated by its expected flow rate which provides scram signal. The SAFETY LIMIT shall be assumed to be 100% core flow at exceeded when scram is 100% power). l accomplished by a means other AW - Dif f erence between than the expected scram two-loop and single-loop signal. effective drive flow at the same core flow.

l y- s,= J s .'. 1

.(.

1.1 Bases

.(Cont'd)

C. Power Transient Plant safety analyses have shown that the scrams caused by exceeding any safety setting will assure that the SAFETY LIMIT of l Specification 1.lA or 1.lB will not be exceeded. Scram times are

-checked periodically to assure the insertion times areLadequate.

The thermal power transient resulting'when a scram is accomplished other than by.the expected scram signal (e.g., scram from neutron flux following closure of the main turbine stop valves) does not necessarily cause fuel damage. However, for this specification a-SAFETY LIMIT violation will be assumed when a - scram is only l~

accomplished by means of a backup feature of the plant design. ' The -

I; concept of not approaching a SAFETY LIMIT provided scram signals are

{~ OPERABLE is supported by the extensive plant safety analysis.

The computer provided with Cooper Nuclear Station has a sequence l annunciation program. which will indicate the sequence in which events such as scram, APRM trip initiation, pressure scram initiation, etc. occur. This program also indicates when the scram setpoint is cleared. This will provide information on how long a scram condition exists and thus provide some measure of the ensrgy added during a transient. Thus, computer information normally will' l be available for' analyzing scrams; however, if - the ' computer information should not be available for any scram analysis, Specification 1.1.C will be relied on to determine if a SAFETY LIMIT l has been violated.

D. Reactor Water Level (Shutdown Condition)

During periods when the reactor is SHUTDOWN, consideration must also l be given to water level requirements due to the effect of decay heat. If reactor water level should drop below the top of . the active fuel during this time, _ the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation. The core can be cooled sufficiently should the water level be reduced to-two-thirds the core height. Establishment -of the SAFETY LIMIT at l 18 inches above the top of the fuel provides adequate margin.

References for 1.1 Bases

1. " General Electric Standard Application for' Reactor Fuel,"

NEDE-240ll-P-A-(latest approved revision),

2. " Cooper Nuclear Station Single-Loop Operation," NEDO-24258, May, 1980.

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l f j i

NOTES ~FOR TABLE 3.1.1 l

1. There shall be two OPERABLE or tripped TRIP SYSTEMS for each function. .If the j minimum number of OPERABLE INSTRUMENT CHANNELS for a TRIP SYSTEM cannot be met, the ,

affected TRIP SYSTEM shall be placed in the safe (tripped) condition, or the appropriate actions listed below shall be taken.

A. Initiate insertion of OPERABLE rods and complete insertion of all OPERABLE rods l l within four hours. I B. Reduce power to less than 30% of RATED POWER.

l l C. Reduce power level to IRM range and place Reactor Mode Selector Switch in the -l STARTUP position within eight hours and depressurize to less than 1000 psig.  ;

9 D. Reduce turbine load and close Main Steam Isolation Valves within eight hours. l c

2. Permissible to bypass, with control rod block, for Reactor Protection System reset' l in REFUEL and SHUTDOWN positions of the Reactor Mode Selector Switch. l

3. This note deleted. I i

~

4. Permissible to bypass when turbine first stage pressure is less than 30% of full '

load.

5. IRMs are bypassed when APRMs are onscale and the Reactor Mode Selector Switch is in l  !

the RUN position. j

6. The design permits closure of any two lines without a full scram being initiated.

7. When the reactor is suberitical, fuel is in the vessel, and the reactor water f temperature is less than 212*F, only the following trip functions need to be

• OPERABLE: l A. Reactor Mode Selector Switch in SHUTDOWN. l

. g f at 120/125 indicated scale.

D. APRM (15%) high flux scram. l

8. Not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.

9. Not required to be OPERABLE while performing low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MW(t).

10. Not required to be OPERABLE when the reactor pressure vessel head is not bolted to -l  !

the vessel.

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COOPER NUCLEAR STATION ~

TABLE 4.2.C -

SURVEILIANCE REQUIREMENTS FOR ROD WITHDRAWAL BLOCK INSTRUMENTATION Functional Function Test Freo. Calibration Freo. Instrument Check APRM Upscale (Flow Bias) (1) (3) Once/3 Months Once/ Day APRM Upscale (Startup Mode) (1) (3) Once/3 Months Once/ Day APRM Downscale (1) (3) Once/3 Months Once/ Day APRM Inoperative (1) (3) N.A. Once/ Day RBM Upscale (Power Referenced) (1) (3) Once/6 Months Once/ Day RBM Power Range (3) Once/6 Months N.A.

RBM Downscale (1) (3) Once/6 Months once/ Day RBM Inoperative (1) (3) N.A. Once/ Day IRM Upscale (1) (2) (3) Once/3 Months Once/ Day IRM Downscale (1) (2) (3) Once/3 Months Once/ Day IRM Detector Not Full In (2) (Once/0PER- Once/0PERATING CYCLE (10) Once/ Day ATING CYCLE)

IRM Inoperative (1) (2) (3) N.A. N.A.

4 Once/ Day y SRM Upscale (1) (2) (3) Once/3 Months SRM Downscale (1) (2) (3) Once/3 Months Once/ Day SRM Detector Not Full In (2) (Once/0PER- Once/0PERATING CYCLE (10) N.A.

ATING CYCLE)

SRM Inoperative (1) (2) (3) N.A. N.A.

Flow Bias Comparator (1) (8) Once/0PERATING CYCLE N.A. l Flow Bias Upscale (1) (8) once/3 Months N.A.

Rod Block Logic (9) N.A. N.A.

SDV High Water Level Quarterly Once/0PERATING CYCLE N.A.

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

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.4' STANDBY LIOUID CONTROL SYSTEM 4.4 STANDBY LIOUID CONTROL SYSTEM Applicability: Apolicabilitv:

-Applies to the operating status of Applies to the surveillance l the Standby Liquid Control (SLC) requirements of the Standby Liquid >

I System. Control (SLC) Syster.

Obiective: Objective:

l -' To assure the OPERABILITY of a To verify the OPERABILITY of the SLC l system with the capability to System.

SHUTDOWN the reactor and maintain the SHUTDOWN condition without the '

use of control rods.

Specification: Specification:

l A. Normal System Operation A. Normal' System Operation l r l 1. During periods when fuel is in the The OPERABILITY of the SLC System reactor and prior to startup from a shall be shown by the performance of l Cold Condition, the Standby Liquid the following tests:

Control System shall be operable, except as specified in 3.4.B below. 1. At least once each 3 months each This system need not be operable subsystem shall be tested for when the reactor is in the Cold OPERABILITY by recirculating Condition and all control rods are demineralized water to the test tank fully inserted and Specification and verifying each pump develops a 3.3.A is met, flow rate 2 38.2 gpm at a discharge pressure 2 1300 psig.

2. At least once during each OPERATING CYCLE:

a. Check that the settings of the subsystem relief valves are 1450 < P < 1680 psig and the valves e will reset at P 2 1300 psig.

b. Manually initiate the system, except explosive valves, and pump boron solution from the SLC Storage Tank l through the recirculation path. I Verify each pump develops a flow rate 2 38.2 gpm at a discharge pressure 2 1300 psig. After pumping boron solution the system will be flushed with demineralized water.

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'# -LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.4 4.4.A.2 (Cont'd.)

c. Manually initiate one of the SLC SystemPumpsandpumpdemineralizedl water into the reactor vessel from the test tank.

These tests check the actuation of the explosive charge of the tested subsystem, proper operation of the valves, and pump OPERABILITY. The replacementchargestobeinstalledl will be selected from the same manufactured batch as a previously.

tested charge.

d. Both subsystems, including both

• cxplosive valves, shall be tested in the course of two OPERATING CYCLES. l B. Operation with Inocerable B. Surveillance with Inoperable Comoonents: Components:

1. From and after the date that one J. When a subsystem is found to be subsystem is made or found to be inoperable, the OPERABLE subsystem inoperable, Specification 3.4.A.1 shall be verified to be OPERABLE shall be considered fulfilled and immediately and daily thereafter continued REACTOR POWER OPERATION until the inoperable subsystem is permitted provided that the OPERABLE returned to an OPERABLE condition. l subsystem remains OPERABLE and the inoperable subsystem is returned to an OPERABLE condition within seven days.

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s, E LIMITING CONDITIONS FOR OPERATION SURVEILIANCE REOUIREME?TTS 3.4.C 4.4.C l

C. Sodium Pentaborate Solution C. Sodium Pentaborate Solution At all times when the SLC System is The following tests shall be required to be OPERABLE the following conditions shall be met: performedtoverifytheavailabilityl of the liquid control solution:

1. The net volume versus concentration 1. Volume: Check and record at least of the liquid control solution in once per day.

the SLC Storage Tank shall be maintained as required in Figure 3.4.1.

2. The temperature of the liquid 2. Temperature: Check and record at control solution shall be maintained least once per day.

above the curve shown in Figure 3.4.2. 3. Concentration: Check and record at least once per month. Also check D. If specification 3.4.A through C concentration anytime water or boron cannot be met, the reactor shall be is added to the solution or solution placed in a Cold Shutdown Condition temperature is below the temperature with all operable control rods fully required in Figure 3.4.2.

inserted within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. l P

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e 3.4 BASES STANDBY LIOUID CONTROL SYSTEM A. The Standby Liquid Control (SLC) System consists of two, distinct subsystems, eachcontainingonepositivedisplacementpumpandindependentsuctionfromthel SLC storage tank, and discharge to a common injection header through parallel explosive valves. The purpose of the SLC System is to provide the capability

'of bringing the reactor from RATED POWER to a cold, xenon-free . SHUTDOWN CONDITION assuming that none of the withdrawn control rods can be inserted.

To meet this objective, the system is designed to inject a quantity of boron that produces a concentration of 660 ppm of boron in the reactor pressure  ;

vessel in less than 125 minutes. The 660 ppm concentration in the reactor pressure vessel is required to bring the reactor from RATED POWER to a 3.0 percent Ak subcritical condition, considering the hot to cold reactivity difference, xenon poisoning, etc. The time requirement for inserting the boron solution was selected to override the rate of reactivity insertion caused by cooldown of the reactor following the xenon poison peak.

The conditions under which the SLC System must provide shutdown capability are identified in Limiting Conditions for Operation. If no more than one OPERABLE control rod is withdrawn, the basic shutdown reactivity requirement for the core is satisfied and the SLC System is not required. Thus, the basic reactivity requirement for the core is the primary determinant of when the SLC System is required. ,

The minimum limitation on the relief valve setting is intended to prevent the recycling of liquid control solution via the lifting of a relief valve at too low a pressure. The upper limit on the relief valve setting provides system protection from overpressure.

B. Only one of the two SLC subsystems is needed for operating the system. One l inoperable subsystem does not immediately th; eaten shutdown capability, and reactor operation can continue while the inoperable subsystem is being repaired. The seven day completion time is based on the availability of an OPERABLE subsystem capabis cf perf:,rming the intended SLC system function and the low probability of a Design Basis Accident (DBA) or severe transient occurring concurrent with the failure of the Control Rod Drive (CRD) system to shut down the plant.

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3.4 BASES (cont'd.)

C. Level indication and alarm indicate whether the solution volume has changed, r which might indicate a solution concentration change. The test interval has been established in consideration of these factors. Temperature and liquid level alarms for the system are annunciated in the control room.

The solution is kept at least 10*F above the saturation temperature to guard against boron precipitation. The margin is included in Figure 3.4.2.

The volume versus concentration requirement of the solution is such that, '

should evaporation occur from any point within the curve, a low level alarm will annunciate before the temperature versus concentration requirements are exceeded.

The quantity of stored boron includes an additional margin (25 percent) beyond the amount needed to shutdown the reactor to allow for possible imperfect ,

mixing of the chemical solution in the reactor water. A minimum quantity of 3132 gallons of solution having a 16.0 weight percent sodium pentaborate concentration, or the equivalent as shown in Figure 3.4.1, is required to meet this shutdown requirement.

The NRC's final rule on Anticipated Transients Without Scram (ATWS),

10CFR50.62, requires that the SLC System be modified to provide a minimum flow l capacity and boron content equivalent in control capacity to 86 gpm of 13 weight percent sodium pentaborate solution for a 251 inch I.D. vessel. For Cooper Nuclear Station, with a 218 inch I.D. vessel, the equivalent minimum ,

flow rate is 66 gpm of 13 weight percent sodium pentaborate. This equivalence is met with both SLC pumps supplying their minimum flow rate of 38.2 gpm with a solution concentration of at least 11.5 weight percent of sodium pentaborate.

Because ATWS is a very low probability event and is considered to be beyond the ,

design basis for CNS, the surveillance and limiting condition for operation requirements need not be more stringent than the original SLC system design basis requirements. The SLC system changes made as a result of the ATWS rule do not invalidate the original system design basis.

4.4 BASES STANDBY LIOUID CONTROL SYSTEM r

Experience with pump OPERABILITY indicates that testing once each three months, in combination with the tests during each OPERATING CYCLE, is sufficient to maintain pump performance. The only practical time to fully test the SLC system is during a REFUELING OUTAGE. Various components of the system are individually tested L periodically, thus making unnecessary more frequent testing of the entire system. j The bases for the surveillance requirements are given in USAR section III-9.6, and the details of the various tests are discussed in section III-9.5. The solution temperature and volume are checked at a frequency to assure a high reliability of operation of the system should it ever be required.

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4565 gol. ,

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(ll.5 % )

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is of, .liI.l.I,l.I,1il,I,I,I,1,1,1,l,I,t,1,I,1,1,1,1,1,i,1,i,1,1,1 2000 2500 3000 3500 4000 4500 5000 l Net Tonk Volume (gal.) (

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NEBRASKA PUBLIC POWER DISTRICT COOPER NUCLEAR STATION 1 1

Sodium Pentaborate Solution Volume-Concentration.

Requirements FIGURE 3.4.1 112

5 -

1 BASES:

3.6.H and 4.6.H I Snubbers l

Snubbers are designed to prevent unrestrained pipe motion under dynamic loads as might I occur during an earthquake or severe transient, while allowing normal thermal motion during STARTUP and SHUTDOWN. The consequence of an inoperable snubber is an increase {  ;

in the probability of structural damage to piping as a result of a seismic or other event initiating dynamic loads. It is therefore required that all snubbers required to protect the primary coolant system or any other safety system or component be OPERABLE during REACTOR POWER OPERATION. l Because the snubber protection is required only during relatively low probability events, a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed for repairs or replacement. Since plant STARTUP should not commence with knowingly defective safety related equipment, Specification 3.6.H.3 prohibits STARTUP with inoperable snubbers.

All safety related snubbers are visually inspected for overall integrity and OPERABILITY. l l The inspection frequency is based upon maintaining a constant level of snubber protection. Thus the required inspection interval varies inversely with the observed snubber failures. The . number of inoperable snubbers found during a required inspection determines the time interval for the next required inspection. Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection. However, the results of such early inspections performed before the original required time interval has elapsed (nominal time less 25%) may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspection interval will override the previous <

schedule.

When the cause of the rejection of a snubber is clearly established and remedied for that snubber and for any other snubbers that may be generically susceptible, and verified by functional testing, that snubber may be exempted from being counted as inoperable. Generically susceptible snubbers are those which are of a specific make or ,

model and have the same design features directly related to rejection of the snubber  !

by visual inspection, or are similarly located or exposed . to the same environmental  ;

conditions, such as temperature, radiation and vibration.  !

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1

• 6 LIST OF NRC COMMITMENTS l ATTACHMENT 3 l Correspondence No: NLS950115 The following table identifies those actions committed to by the District in this document. Any other actions discussed in the submittal represent intended or planned actions by the District. They are described to the NRC for the NRC's information and are not regulatory commitments. Please notify the Licensing Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.

COMMITTED DATE COMMITMENT OR OUTAGE None l PROCEDURE NUMBER 0.42 l REVISION NUMBER 0 l PAGE 10 OF 16 l l l

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