ML20072M621

Confirmatory Order Verifying 801216,810513,820111 & 830512 Commitments to Install Permanent Scram Discharge Sys Mods Prior to Cycle 6 Operation

ML20072M621
Person / Time
Site:	FitzPatrick
Issue date:	06/24/1983
From:	Eisenhut D Office of Nuclear Reactor Regulation
To:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
Shared Package
ML20072M626	List: ML20072M621
References
NUDOCS 8307140509
Download: ML20072M621 (5)

Text

.0,

?!

4 7590-01 i

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 1

In the Matter of

)

(POWER AUTHORITY OF THE STATE Docket No. 50-333 0F NEW YORK)

(James A. FitzPatrick Nuclear Power Plant)

)

CONFIRMATORY ORDER I.

The Power Authority of the State of New York (the licensee) is the holder of Facility Operating License No. DPR-59 which authorizes the licensee to operate the James A. FitzPatrick Nuclear Power Plant (the facility),

at power levels not in excess of 2436 megawatts thermal. The facility is a boiling water reactor located at the licensee's site in Oswego County, New York.

II.

During a routine shutdown of Browns Ferry Unit No. 3 on June 28, 1980, 76 of 185 control rods failed to fully insert in response to a manual scram from approximately 30% power. All rods were subsequently inserted within 15 minutes and no reactor damage or hazard to the public occurred.

However, the event did cause an in-depth review of the current BWR Control Rod Drive Systems which identified design deficiencies requiring both short and long-tenn corrective measures.

These measures were set forth in the " Generic Safety Evaluation j

Report BWR Scram Discharge System", dated December 1,1980, prepared by the NRC staff.

8307140509 830624 PDR ADOCK 05000333 j

p PDR

)

I.' -

I

.gi

(

7590-01 l

4 2

To provide reasonable assurance of safe operation pending implementation of long-term corrective measures, the short-term corrective measures have been implemented by IE Bulletin 80-17 (with supplements) and Orders issued on January 9,1981.

The Generic Safety Evaluation Report (SER) dated Decenber 1,1980, endorsed the criteria and technical bases that were developed by a BWR Owners Subgroup for use in implementing permanent system modifications to correct identified deficiencies. These criteria were designated as either functional, safety, operating, design, or surveillance, and when taken as a whole,. comprise an adequate set of criteria to resolve the issues raised during the Browns Ferry event investigation.

The SER further described an acceptable means of compliance with each criterion. Pre-implementation approval of permanent modifications using the methods described in the SER for compliance with the criteria will not be required. Alternate methods of compliance will require specific NRC approval in advance of implementation.

In addition to the criteria proposed by the BWR Owners Subgroup, the SER added a criterion to address the potential for common cause failures of the scram level instrumentation. An acceptable means of complying with this criterion was the addition of diversity in the design.

The addition of diverse instrumentation on the Scram Discharge Instrumented i

Volume will minimize recurrence of known common cause failures and thus improve system reliability.

i

.[..

y 4

\\

7590-01 1

l t '

Therefore, we have concluded that diverse instrumentation should be provided as required in the SER, with one exception:

Alternat.ve 2(d)(ii) has been deleted as a possible means of providing diversity, due to its reliance on prompt operator action. The use of level sensors employing different operating principles, or the use of level sensors made by a different manufacturer, continues to be acceptable means of providing diverse instrumentation.

On October 1,1980 letters were sent to all BWR licensees requesting a commitment to reevaluate the present scram system and modify it as necessary to meet the design and performance criteria developed by the BWR Owners Subgroup. The letter also requested a schedule for implementation.

III.

Because the implementation of modifications to meet the criteria proposed by the BWR Owners Subgroup and endorsed by the NRC staff will restore the margins of safety in the BWR scram system, we have determined that these. modifications should be completed on an expeditious schedule.

In response to our letter of October 1,1980 and additional discussions with the NRC staff, the licensee committed, by letters dated December 16, 1980 and supplemented by letters dated May 13, 1981, January 11, 1982 and May 12, 1983 to install the long term modifications before recetor operation in Cycle 6.

In view of the foregoing, we have determined that these commitments ire required in the interest of public health and safety and should,- therefore, be confirmed by an immediately effective order.

I

e

~

a 4

7590-01 4-IV.

Accordingly, pursuant to sections 103, 1611, and 182 of the Atomic Energy Act of 1954, as amended, and the Commission's regulations in 10 CFR Parts 2 and 50, IT IS HEREBY ORDERED EFFECTIVE IMMEDIATELY THAT:

1.

The licensee shall install the lor.g term BWR scram discharge system modifications in conformance with the staff's Generic SER, which incorporates the BWR Owners Subgroup criteria, before reactor operation in Cycle 6 or, in the alternative, the licensee shall place and maintain tne facility in a cold shutdown or refueling mode of operation until such modifications are made. Extensions of time for installation may be granted for good c&use shown by the licensee. The modifications shall include diverse instrumentation as provided in the SER with the I

exception that alternative 2(d)(ii) will not be accepted.

2.

For those cases in which a different method of complying with the criteria than that described in the SER is chosen, the licensee shall submit the design details and supporting analyses for approval to the Director, Division of Licensing, Washington, D. C. 20555 with a copy to the Regional Administrator of the appropriate NRC regional office, at least 3 months prior to the required implementation date.

3.

Technical Specification changes required for operation with the niodified system shall be submitted at least 3 months prior to the required

~

implementation date or within 30 days'of-the-date of this Order, whichever l

l is-later.

e M

. y 7 xv

'4

.?

e

..i.,

?

c.'

7590-01 s

c x

3

\\

y

'\\

,q 3

V.

t

s. t.

The licensee may request a hearing on this Order within 25-days of the

'x.

date of publication of this Order in the F_ederal Register. -A, request for hearing shall be submittea to' the Df rector,'\\ = ision of Licensing, Offic.e Div s

-s x,

of Nuclear Reactor Re'quh, tion, U.S. Nuclear ReOulatory Commichien. s\\

t -

s t 4,

~,

Washington, D.C.

2 $55. ( A copy of the request shall also be sent to the h.

Executive legal DiriE:to? at the U.S. Nuclear Regulatory Comnfission, c

Washington, D.C.

'20555. A REQUEST FOR HEAk:NG SHALL NOT STAY THE'.

y IMMEDIATE EFFECTIVENESS OF THIS ORDERI

~

~

If a hearing is requested by the licensee, the Commissibn'will issue

~

an order designating the time and place of any such hearing'.

If a hearing 4

s..

is held, the issua to be considered at sucit a heyring shallbe whether the

c a

licensee should comply with the conditionc Tet for.h ir. Section IV~.

t'

~

?

of this Order.

s'

s.,

s e

e The request for infomation made in this Or' der was! approved by OMB

~

~

s

- x.. N under clearance number 3160-0083 which ~ expires $. n December 31, 1983. Comments N ~'

s e -

.s s

s.

on burden and duplication may be directed to the Office,cf Management aqd e

Budget, Reports Management [ Room 3208, New Executive Of fice Building,

/

Washington, D.C.

N,

._ s

.s s

+ -

m m

This Order is effective upon issuanca.

v FOR THE NUCLEAR REGULATORY COMMISSION -

\\'

s,,

s.

C

,LL Q %,i d, j 3 ~ Q,7,

~

g

f s s s

t

-(

Darrell G jisenhut, Director I,

l'

- -Division of LicensirN 1

' ',: s Office of Nuclear Reacter Regulation

.N Dated at Bethesda, Maryland 1

k this 24th day of. June 1983.

(

^

4 i

s

\\

k

.y x,

~

SAFETY LIM,ITS AND LXMITING SAFETY SYSTEM SETTINGS e

2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor protection system instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2.1-1.

APPLICABILITY:

As shown in Table 3.3.1-1.

ACTION:

With a reactor protection system instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2.1-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its setpoint adjusted consistent with the Trip Setpoint value.

9

)

k s

GE-STS (BWR/4) 2-3 ao m.,

ra 5x op,-

g TABLE 2.2.1-1 l

.i REAC10R PROTECTION SYSTEM INSTRUMENTATION SETrotNIS

.2 ALLOWADLE

.i O FUNCTIONAL UNIT TRIP SETPOINT VALUES

c j

R 1.

Intermediate Range Monitor, Neutron Flux-liigh 1

O

-< (120)/(125) divisions 1 (122)/(125) divistoi of full scale of full scale 2.

Average Power Range Monitor:

a.

Neutron Flux-Upscale, Setdown 1 (15)% of RATED THERMAL POWER 1 (20)% of RATED.

l TilERMAL POWER i

b.

Flow lif ased Simulated Thermal Power-Upscale

1) Flow Hiased 1 0.66 Wt(51)%, with i 0.6G W+(54)%, with a maximum of a maximim of r

2) liigh Flow Clamped 1 (113.5)% of RATED i (115.5)% of RATED 4

THERMAL POWER TilERMAL POWER c.

Fixed Neutron Flux-Upscale 1 (118)% of RATED THERMAL POWER 1 (120)% of RATED TilERMAL POWER

}

i d.

Inoperative NA NA j

(c.

Downscale 1 (5)% of RAIED 1 (3)% of RATED IllERMAL POWER TilERMAL POWER)

[

3.

Reactor Vessel Steam Dome Pressure - liigh i (1045) psis; i (1065) psig

~

4.

Reactor Ver.sel Water Level - Low, Level 3 1 (12.5) inches above instrudent 1 (11.0) inches above zero*

Instrament zero 5.

Main Steam Line Isolation Valve - Closure 1 (6)% closed 1 (71% closed i

6.

Main Steam Line Radiation -

liigh 5 (2.5) x full power background

$ (3.0) x full power

< s hacknround 7.

(Primary Containment) (Drywell) Pressure - High i (1.69) psig 1 (1.89) psig 8.

Scram Dir.ch.irge Volume Water Level - liigh i (36)% of full scalo 1 (39)% of full scale i l

9.

Turbine Stop Valve - Closure 1 (5)% closed 1 (7)% closed i

10.

Turbine Control Valve Fast Closure, i

Trip oil Pressure - Low i

1 (500) psig 1(

) p>ig 11.

Reactor Mode Switch Shutdown Position NA NA-12.

Manual Scram NA NA i

6' "See Bases Iigure U 3/4 3-1.

l,

.?

LIMfTfNG SAFETY SYSTEM SETTING BASES i

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) 8.

Scram Discharge Volume Water Level-High The scram discharge volume receives the water displaced by the motion of

~

the control rod drive pistons during a reactor scram.

Should this volume fill up to a point where there is insufficient volume to accept the displaced water at pressures below 65 psig, control rod insertion would be hindered.

The reac-tor is therefore tripped when the water level has reached a point high enough to indicate that it is indeed filling up, but the volume is still great enough to accommodate the water from the movement of the rods at pressures below 65 psig when they are tripped.

The trip setpoint for each scram discharge volume is equivalent to a contained volume of (

) gallons of water.

9.

Turbine Stop Valve-Closure The turbine stop valve closure trip antfcipates the pressure, neutran flux, and heat flux increases that would result from closure of the stop valves.

With a trip setting of (5)% of valve closure from full open, the l

resultant iherease in heat flux is such that adequate thermal margins are maintained during the worst case transient (assuming the turbine bypass valves (fail to) operate).

l 10.

Turbine Control Valve Fast Closurs, Trip Oil Pressure-Low The turbine control valve fast closure trip anticipates the pressure, neutron flux, and heat flux increase that could result from fast closure of the turbine control valves due to load rejection coincident with~ failure of the turbine bypass valves.

The Reactor Protection System initiates a trip when fast closure of the control valves is initiated by the fast acting sole-noid valves and in less than (30) milliseconds after the start of control valve fast closure.

This is achieved by the action of the fast acting solenoid valves in rapidly reducing hydraulic trip oil pressure at the main turbine control valve actuator disc dump. valves.

This loss of'pressur'e is sensed by press'0re

~

switches whose epntacts form the one-out-of-two-twice logic input to the Reactor Prctection System.

This trip setting, a faster closure time, and a different valva characteristic from that of the turbine stop valve, combine to produce transients which are very similar to that for the stop valve.

Relevant tran-sient analyses are discussed in Section (15.1.0) of the Final Safety Analysis Report.

11.

Reactor Mode Switch Shutdown Position The reactor mode switch Shutdown position is a redundant channel to the automatic protective instrumentation channels and provides additional manual reactor trip capability.

'12.

Manual Scram The Manual Scram is a redundant channel to the automatic protective instrumentation channels and provide,s manual reactor trip capability.

GE-STS (SWR /4)

B 2-9

.1

~

REACTIVITY CONTROL SYSTEMS 3/4.1.3 CONTROL RODS CONTROL R00 OPERABILITY LIMITING CONDITION FOR OPERATION 3.1.3.1 All control rods shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With one control rod, inoperable due to being immovable, as a t esult of a.

excessive friction or mechanical interfe~rence, or known to be untrippable:

1.

Within one hour:

a)

Verify that the inoperable control rod, if withdrawn, is separated from all dther inoperable control rods by at least two control cells in all directions.

b)

Disarm the associated directional control valves ** either:

l

1) Electrically, or

2) Hydraulically by closing the drive water and exhaust water isolation valves.

c)

Comply with Surveillance Requirement 4.1.1.c.

Otherwise, be in at least HOT SHUTDOWN pithin the next 12 hcurs.

l 2.

Restore the inoperable control rod to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> l

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.

With one or more control rods trippable but inoperable for causes other l

than addressed in ACTION a, above:

1.

If the inoperable control rod (s) is withdrawn, within one hour:

a)

Verify that the inoperable withdrawn control rod (s) is separated from all other inoperable control rods by at least two control cells in all directions, and b)

Demonstrate the insertion capability of the inoperable withdrawn l

control rod (s) by inserting the control rod (s) at least'one notch by drive water pressure within the normal operating range".

Otherwise, insert the inoperable withdrawn control rod (s) and disarm the associated directional control valves ** either:

a)

Electrically, or 1

b)

Hydraulically by closing the drive water and exhaust water l

isolation valves.

"The inoperable control rod may then be withdrawn to a position no further withdrawn than its position when found to be inoperable.

• • May be rearmed i.ntermittently, under administrative control, to permit testing associated with restoring the control rod to OPERABLE status.

GE-STS (BWR/4) 3/4 1-3

/

O react 1VITYCONTROLSYSTEMSI QMITINGCONDITIONFOROPERATION(Continued)

ACTION (Continued) 2.

If the inoperable control rod (s) is inserted, within one hour disarm the associated directional control valves ** either:

a)

Electrically, or b)

Hydraulically t;y closing the drive water and exhaust water isolation valves.

l Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

l With more than 8 control rods inoperable, be in at least HOT SHUTDOWN c.

within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE REOUIREMENTS 4.1.3.1.1 The scram discharge volume drain and vent valves shall be demonstrated OPERABLE bf:

i I

At least once per 31 days verifying each valve to be open,* and l

a.

b.

At least once per 92 days cycling each, valve through at least one l

complete cycle of full travel.

4.1.3.1.2 When above the (preset power level) (low power setpoint) of the RMi l

and RSCS, all withdrawn control rods not required to have their directional control valves disarmed electrically or hydraulically shall be demonstrated OPERABLE by moving each control rod at least one notch:

At least once per 7 days, and a.

b.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when any control rod is immovable as a result of excessive friction or mechanical interference.

4.1.3.1.3 All control rods shall be demonstrated OPERABLE by performance of Surveillance Requirements 4,1. 3.,2, 4.1. 3. 4, 4.1. 3. 5, 4,1. 3. 6 and 4.1. 3. 7.

"These valves may be closed intermittently for testing under administrative controls.

"May be rearmed intermittently, under administrative control, to permit testing associated with restoring the control rod to OPERABLE status, i

GE-STS (BWR/4) 3/4 1-4

': ?

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.1.3.1.4 The scram discharge volume shall be determined OPERABLE by demonstrating:

The scram discharge volume drain and vent valves OPERABLE, when a.

control rods are scram tested from a normal control rod configura-i tion of less than or equal to (50)% ROD DENSITY at least once per 18 months, by verifying that the drain and vent valves:

1.

Clcse within (30) seconds after receipt of a signal for control rods to scram, and 2.

Open when the scram signal is reset.

l b.

Proper (float) (level sensor) response by performance of a CHANNEL l

FUNCTIONAL TEST of the scram discharge volume scram and control rod block level instrumentation (AT level measuring system) (after each scram from a pressurized condition) (at least once per.31 days).

pm O

.i i

9 j

e e

I e

4 GE-STS (BWR/4) 3/4 1-5

/

=. _.

---

l

,g C.

./

s m

REACTIVITY CONTROL SYSTEMS R00 BLOCK MONITOR LIMITING CONDITION FOR OPERATION 3.1.4.3 Both rod block monitor (RBM) channels shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than r

or equal to (30)% of RATED THERMAL POWER.

l ACTION:

With one RBM channel inoperable, restore the inoperable RBM channel a.

to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and verify that the reactor is not operating on a LIMITING CONTROL ROD PATTERN; otherwise, place the inoperbie rod block monitor channel in the tripped condition within the next hour.

' b.

With both RBM channels inoperable, place at least one inoperable red block monitor channel in the tripped condition within one hour.

SURVEILLANCE REQUIREMENTS 4.1.4.3 Each of the above required RBM channels shall be demonstrated OPERABLE by performance of a:

CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION at the frequencies a.

and for the OPERATIONAL CONDITIONS specified in Table 4.3.6-1.

b.

CHANNEL FUNCTIONAL TEST prior to controi~ rod withdrawal when the reactor is operating on a LIMITING CONTROL R0D PATTERN.

)

GE-STS (BWR/4) 3/4 1-18 4

~

REACTIVITY CONTROL SYSTEMS BASES 3/4.1.3 CONTROL R005

~

The specification of this section ensure that (1) the minimum SHUTDOWN MARGIN is maintained, (2) the control rod insertion times are consistent with those used in the accident analysis, and (3) limit the potential effects of the rod drop accident.

The ACTION statements permit variations from the basic requirements but at the same time impose more restrictive criteria for continued

' operation.

A limitation on inoperable rods is set such that the resultant effect on total rod worth and scram shape will be kept to a minimum.

The requirements for the various scram time measurements ensure that any indication of systematic, problems with rod drives will be investigated on a timely basis.

Damage within the control rod drive mechanism could be a generic problem, therefore with a control rod immovable because of excessive friction or mechanical interference, operation of the reactor is limited to a time period which is reasonable to determine the cause of the inoperability.and at the same time prevent operation with a large number of inoperable control rods.

' Control rods that are inoperable for other reasons are permitted to be taken out of service provided that those.in the nonfully-inserted position are

~

consistent with the SHUTDOWN MARGIN requirements.

The number'of control rods permitted to be inoperable could be more than the eight allpwed by the specification, but the occurrence of eight inoperable rods could be indicative of a generic problem and the reactor must be shutdown j

for investigation and resolution of the problem.

The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent the MCPR from becoming less than (1.06) during the limiting power transient analyzed in Section (15.

) of.the FSAR.

This analysis shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the specifications, provide the required protection and MCPR remains greater than (1.06).

The occurrence of scram times longer then those specified should be viewed as an indication of a systemic problem with the rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reactor for long. periods of time with a potentially serious problem.

The scram discharge volume is required to be OPERABLE so that it will be available when needed to accept discharge water from the control rods during a l

reactor scram and will isolate the reactor coolant system from the containment when required.

Control rods with inoperable accumulators are declared inoperable and Specification 3.1.3.1 then applies.

This prevents a pattern of inoperable i

accumulators that would result in less reactivity insertion on a scram than has been analyzed even though control rods with inoperable accumulators may still be inserted with normal drive water pressure.

Operability of the accumulator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of the reactor.

l GE-STS (BWR/4)

B 3/4 1-2

~

i

.x REACTIVITY CONTROL SYSTDQ BASES CONTROL RODS (Continued)

Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR.

The overtravel position feature provides the only positive means of determining that a red is properly coupled and therefore this check must be performed prior to achieving criticality after completing CORE ALTERATIONS that could have affected the control rod coupling integrity.

The subsequent check is performed as a backup to the initial demon-stration.

In order to ensure that the control rod patterns can be followed and there-fore that other parameters are within their limits, the control rod position indicat'on system must be OPERABLE.

The control rod housing support restricts the outward movement of a control red to less than (3) inches in the event of a housing failure.

The amount of rod reactivity which could be added by this small amount of rod withdrawal is less'than a normal withdrawal increment and will not contribute to any damage to the primary coolant system. The support is not required when there is no pressure to act as a dr.iving force to rapidly eject a drive housing.

The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components. ~

3/4.1.4 CONTROL ROD PROGRAM CONTROLS Control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to result in a peak fuel enthalpy greater than 280 cal /gm in the event of a control rod drop accident.

The specified requences are characterized by homogeneous, scattered patterns of control rod withdrawal. When THERMAL POWER is greater than (20)% of RATED THERMAL POWER, there is no possible rod worth which, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal /gm. Thus requiring the RSCS and RWM to be OPERABLE when THERMAL POWER is less than or equal to (20)% of RATED THERMAL POWER providEs adequate control.

The RSCS and RWM provide automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted.

1he analysis of the rod drop accident is presented in Section (15.

) 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 RBM is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of nigh power density during high pou r operation.

Two channels are provided.

Tripping one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage.

This system backs up the written sequence used by the operator for withdrawal of control rods.

GE-STS (BWR/4)

B 3/4 1-3

. = _ _ -

3/4.3 INSTRUMENTATION 3/4.3.1 REACTLR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation channels 1

shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2.

APPLICABILITY:

As shown in Table 3.3.1-1.

ACTION:

With the number of OPERABLE channels less than required by the Minimum a.

OPERABLE Channels per Trip System requirement for one trip system, place the inoperable channel (s) and/or that trip system in the tripped condi-tion

• within one hour.

The provisions of Specification 3.0.4 are not applicable.

b.

With the number of OPERABLE channels less than required by th'e Minimum OPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system ** in the tripped condition within one hour and take the ACTION required by Table 3.3.1-1.

1 j

SURVEILLANCE REOUIREMENTS 4.3.1.1 Each reactor protection system instrumentation channel 'shall be 4

demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1-1.

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

f 4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE' TIME of each reactor trip functional unit shown in Table 3.3.1-2 shall be demonstrated to be within its limit at least once per 18 months.

Each test shall include at least one channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip system.

l "An inoperable channel need not be placed in the tripped condition where this l

would cause the Trip Function to occur.

In these cases, the incperable channel shall be restored to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or the ACTICN required by Table 3 3.1-1 for that Trip Function shall be taken.

l l

• • If more channels are inoperable in one trip system than in the other, place the trip system with more inoperable channels in the tripped condition, j

except when this would cause the Trip Function to occur.

l l

GE-STS (BWR/4) 3/4 3-1 w. -- _. _ _ _. _ _

TABLE 3.3.1-1 u,

's d

REACTOR PROTECTION SYSTEM INSTRUMENTATION G

e 4

APPLICABLE MINIMUM 4,

8 OPERATIONAL OPERABLE CllANNELS FUNCTIONAL UNIT CONDITIONS PER TRIP SYSTEM (a)

ACTION 1.

Intermediale Range Monitors (b);

a.

Neutron Flux - liigli 2

3 1

(c) 2(d) b.

Inoperative 2

3 1

3, 4 2

2 5

3(d) 3 3

I 2.

Average Power Range Monitor "):

a.

Neutron Flux - Upscale, Setdown 2

2 1

{

3(c) 2(d) 2 i'

b.

Flow Biased Sin.ulated Thermal l

~

Power,- Upscale 1

2 4

c.

Fixed Neutron Flux - Upscale 1

2 4

u d.

Inoperative 1, 2 2

1 3(c) 2(d) 2 l IU) 2

4) l

\\

(e.

Downscale I

3.

Reactor Vessel Steam Dome Pressure - liigh 1,2(I)

I

~

2 1

4.

Reactor Vessel Water Level - Low; Level 3 1, 2 2

1 5.

Main Steam Line Isolation Valve -

Closure I(9) 4 4

't TABLE 3.3.1-1 (Continued)

^

REACTOR PROTECTION SYSTEM INSTRUMENTATION Y

h APPLICABLE MINIMUM R

OPERATIONAL OPERABLE CilANNELS O

FUNCTIONAL UNIT CONDITIONS PER TRIP SYSTEM (a)

ACTION 6.

Main Steam Line Radiation -

liigh 1, 2(I) 2 5

7.

(Primary Containment) (Drywell)

Pressure - liigh 1, 2(h) 2 1

8.

Scram Discharge Volume Water 1

2 1

Level - lli9 1, 2(3) 5 2

3 9.

Turbine Stop Valve - Closure 1(I}

4( }

6

{

Y 10.

Turbine Control Valve Fast Closure, I (9))(I) 2(k) 6 I

Valve Trip System oil Pressure - Low l

11.

Reactor Mode Switch Shutdown-l Position 1, 2 1

1 1

3, 4 1

7 l

l 5

1 3

l

]

12.

Manual Scram 1, 2.

2 1

l 8

l 3, 4 2

l 5

2 9

l l

\\

I A

f

I,'.!

r,;-

. /;,,

-.;t C

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION ACTION 1 Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 2 Verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the Shutdown position within one hour.

ACTION 3 Suspend all operations involving CORE ALTERATIONS

• and insert all insertable control rods within one hour.

ACTION 4 Be in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

1 ACTION 5 Se in STARTUP with the main steam line isolation valves closed

'l within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 6 Initiate a reduction in THERMAL POWER within 15 minutes and reduce turbine first stage pressure to 5 (250) psig, equivalent to THERMAL POWER less than (30)% of RATED THERMAL POWER, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

ACTION 7 Verify all insertable control rods to be inserted within one hour.

ACTION 8 Lock the reactor mode switch in the Shutdown position within one hour.

ACTION 9 Suspend all operations involving CORE ALTERATIONS *, and insert all insertable control rods and lock the reactor mooe switch in the SHUTDOWN position within one hour.

"Except movement of IRM, <5RM or special ~ movable detectors, or replacement of LPRM strings provided SRM instre entation is GPERASLE per Specification 3.9.2.

GE-STS (BWR/4) 3/4 3-4

-d e

n s.

a e

TABLE 3.3.1-1 (Continued) 1 REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a) A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillance without placing the trip system in the tripped i

condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.

(b) This function shall be automatically bypassed when the reactor mode switch is in the Run position.

1 (c) The " shorting links" shall be removed from the RPS circuitry prior to 4

and during the time any control rod is withdrawn

• and shutdown margin demonstrations performed per Specification 3.10.3.

(d) The non-coincident NMS reactor trip function logic is such that all channels go to both trip systems.

Therefore, when the " shorting links" are removed, the Minimum GPERABLE Channels Per Trip System is 4 APRMS and 6 IRMS.

~

(e) An APRM channel is inoperable if there are less than 2 LPRM inputs per level or less than (11) LPRM inputs to an APRM channel.

~

(f) This function is not required to be OPERABLE when the reactor pressure vessel head is unbolted or removed per Specification 3.10.1.

(g) This function shall be automatically bypassed when the reactor, mode switch is not in the Run position.

(h) This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.

(i) With any control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

(j) This function shall te automatically bypassed when turbine first stage pressure is 5 (250) psig, equivalent to THERMAL POWER less than (30)%

of RATED THERMAL POWER.

(k) Aiso actuates the EOC-RPT system.

"No required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.

i l

i l

GE-STS (BWR/4) 3/4 3-5 l

/

o TABLE 3.3.1-2 S

J, REACTOR PROTECTION SYSTEM RESPONSE TIMES TA m

E!

HESl'ONSE TlHE

{

l'UtlCil0ilAL lill_l_l,

_ Qecondsj_

v 1.

Intermediate Range Monitors:

a.

lieutron Flux - liigh NA b.

Inoperative NA il 2.

Average Power Range Monitor *:

a.

Neution Flux - Upscale, Setdown i

NA b.

Flow Biased Simulated Thermal Power - Upscale 5 (0.09)f**)

c.

Fixed Neutron Flux - Upscale 5 (0.09) d.

Inoperative NA (c.

Downscale NA) l R'

3.

Reactor Vessel Steam Dome Pressure - liigh 5 (0.55) 4.

Reactor Vessel Water Level - Low, Level 3

< (1.05)

Y 5.

Main Steam Line Isolation Valve - Closure 3(0.06) 6.

Main Steam Line Radiation - liigh NA j

7.

(Primary Containment) (Drywell) Pressure - liigh NA 8.

Scram Olscharge Volume Water Level - liigh NA 9.

Turbine Stop Valve - Closure '

5 (0.06) 10.

Turbine Control Valve Fast Closure, Irip Oil Pressure - Low

< (0.08)#

11.

Reactor Mode Switch Shutdown Position HA 12.

Manual Scram NA d.

i 4

^ Neutron detectors are exempt from response time testing.

Response time shall be measured from the detector output or from the input of the first electronic component in the channel.

(This provision is not applicable to Construction Permits docketed after January 1, 1978.

See Regulatory Guide 1.18, November 1977.)

• • (Not) Including simulated thermal power time constant, 6.1 1 seconds.

1. Heasured from start of turbine control valve fast closure.

I 4

9 TABLE 4.3.1.1-1 y

REAC10R PROTECTION SYSTEM INSTRUMEtlTATION SURVEILLANCE REQUIREMENTS.

iii CilANNEL OPERATIONAL ~

M CllANNEL 2

furtCTI0tlAL llNIT CllECK

' FUNCTIONAL CllAt!NEL CONDITIONS FOR WillCil TEST CALIBRATION (8)

SURVEILLANCE REQUIRED 1.

Intermediate Range Monitors:

a.

Neutron Flux - liigh S/U,S,(b)

S/U(c);W R

2 l

5 W

R 3,4,5

(

b.

Inoperative ilA W

NA 2,3,4,S 2.

Average Power Range MonitorN):

a.

Neutron Flux -

S/U S,(b)

S/UIC),W SA

?.

Upscale, Setdown S

.W SA 3, 5 b.

Flow Biased Simulated 5,D (U))

5/U(c),y g(d)(e) SA,(R(h))

j I

Thermal Power - Upscale w1 c.

Fixed Neutron Flux -

Upscale S

S/U(c),W W(d), SA 1

m d.

Inoperative NA W

NA 1, 2, 3, 5 l (e.

Downscale S

W' SA 1)

-l ss 3.

Reactor Vessel Steam Dome i

Pressure - liigh (S)

M (R) 1, 2 4.

Reactor Vessel Water Level -

Low, Level 3 (S)

M (R) 1, 2 5.

Main Steam Line Isolation Valve - Closure NA

'M R

1 6.

Main Steam Line Radiation -

liiuh S

H R

1, 2(I) j 7.

(Primary Containment) (Drywell) l Pressure - liigh (S)

M (R) 1, 2 0

4

TAllLE 4.3.1.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS OPERATIONAL CllANNEL

^

E CllANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WilICil

{

FUNCTIONAL UNIT CNECK TEST CALIBRATION SURVEILLANCE REQUIRED v

8.

Scram Discharge Volume, Water l

Level - High (S)

M (R) 1, 2, 5(j)

,1, ~

9.

Turbine Stop Valve - Closure (S)

M (R) 1 10.

Turbine Control Valve Fasb I

Closure Valve Trip System Oil Pressure - Low (S)

M (R) 1 11.

Reactor Mode Switch Shutdown Position NA R

NA 1,2,3,4,5

,s*

Y 1,2,3,4,S 12.

Manual Scram NA M

NA oo (a) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(b) The IRM and SRM channels shall be determined to overlap for at least (h) decades during each startup i

after entering OPERATIONAL CONDITION 2 and the IBM and APRM channels shall be determined to overlap for at 1 cast (4) decades during each controlled shutdown, if not performed within the previous 7 days.

(c) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to startup, if not performed within the previous 7 days.

(d) This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER > 25% of RATED 3'

TilERMAL POWER.

Adjust the APRM channel if the absolute difference is greater than 2% of RATED TilERMAL POWER.

Any APRM channel gain adjustment made in compliance with Specification 3.2.2 shall not be j

included in determining the absolute difference.

(e) This calibration shall consist 6f the adjustment of the APRM flow biased channel to conform to a 4

calibrated flow signal.

(f) The LPRMs, shall be calibrated at least once per 1000 effective full power hours (EFPH) using.the TIP system.

(0) Verify measured core flow to be greater than or equal to established core flow at the existing pump speed.

((h) This calibration shall cons'ist of (the adjustment, as required, of) (verifying) the 6 i l second simulated thermal power time constant. )

(i) This function is not required to be OPERABLE when the reactor pressure vessel head is unbolted or removed per Specification 3.10.1.

(j) With any control rod withdrawn..Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

i i.')

INSTRUriENTATION 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.6.

The control rod block instrumentation channels shown in Table 3.3.6-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.6-2.

APPLICABILITY: As shown in Table 3.3.6-1.

ACTION:

With a control rod block instrumentation channel trip setpoint less a.

conservative than the value shown in the Allowable Values column of Table 3.3.6-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

b.

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channe.ls per Trip Function requirement, take the ACTION required by Table 3.3.6-1.

SURVEILLANCE REQUIREMENTS 4.3.6 Each of the above required control rod block trip systems and instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.6-1.

GE-STS (BWR/4) 3/4 3-47

TABLE 3.3.6-1 S

CONTROL R0D BLOCK INSTRUMENTATION "4

MINIMUM APPLICABLE t '.

OPERABLE CilANNELS OPERATIONAL G

TRIP FUNCTION PER TRIP FUNCTION CONDITIONS ACTION k

1.

ROD DLOCK MONITOR (a)

O a.

Upscale 2

1*

60 b.

Inoperative 2

1*

60 c.

Downscale 2

1*

60 2.

APRM a.

Flow Biased Neutron Flux -

~

g 1

61 Upscale 4

b.

Inoperative 4

1, 2, 5 61 c.

Downscale 4

1 61, d.

Heutron Flux - Upscale, Startup 4

2, 5 61 3.

50llRCE RANGE MONITORS R'

Detector not full in(b) 3 2

61 a.

2 5

61 3

2 61 Upscale (c) b.

2 5

61 ca 3

2

(

Inoperative (c) c.

d.

Downscale(d) 3-2 6

4.

INTERMEDIATE RANGE MONITORS II'))

a.

Detector not full in 6

2, 5 61

N.

b.

Upscale 6

2, 5 61 4

Inoperati 6

2, 5 61 j

c.

Downscale{g) 6 2, 5 61 d.

5.

SCRAM DISCllARGE VOLUME a.

Water Level-liigh (2) 1, 2, 5**

62 b.

Scram Trip Bypass (2)

(1, 2,) 5**

62 6.

REACTOR COOLANT SYSTEM RECIRCULATION FLOW a.

Upscale 2

1 62 b.

Inoperative 2

1 62 c.

(Comparator) (Downscale) 2 1

62

TABLE 3.3.6-1 (Continued)

CONTROL ROD BLOCK INSTRUMENTATION ACTION ACTION 60 Declare the RBM inoperable and take the ACTION required by Specification 3.1.4.3.

ACTION 61 With the number of OPERABLE Channels:

a.

One less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channelto OPERABLE status within 7 days or place the inoperable channel in the tripped condition within the next hour.

b.

Two or more less than required by the Minimum OPERASLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within one-hour.

ACTION 62 With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within one hour.

NOTES With THERMAL POWER > (30)% of RATED THERMAL POWER.

With more than une control rod withdrawn. ' Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

a.

The RBM shall be automatically bypassed when a peripheral control rod is selected (or the reference APRM channel indicates less than (30)% of _

RATED THERMAL POWER).

b.

This function shall be automatically bypassed if detector count rate is

> 100 cps or the IRM channels are on range (3) or higher.

c.

This function shall be automatically bypassed when the associated IRM channels are on range 8 or higher.

d.

This function shall be automatically bypassed when the IRM channels are on range 3 or higher.

e.

This function shall be automatically bypassed when the IRM channels are on range 1.

l I

l GE-STS (BWR/4) 3/4 3-49 l

TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS d

TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE 1.

R00 BLOCK MONITOR R

a.

Upscale

< 0.66 W + (40)%

< 0.66 W + (43)%

^

8 b.

Inoperative NA 5A c.

Downscale 1 (5)% of RATED THERMAL POWER 1 (3)% of RATED THERMAL POWER 2.

APRM a.

Flow Biased Neutron Flux -

Upscale 5 0.66 W + (42)%*

5 0.66 W + (45)%"

b.

Inoperative NA HA c.

Downscale 1 (5)% of RATED THERMAL POWER 1 (3)% of RATED THERMAL POWER d.

Neutron Flux - Upscale, Startup 5 (12)% of RATED THERMAL POWER

$ (14)% of RATED THERMAL POWER 3.

SOURCE RANGE MONITORS l

a.

Detector not full in NA NA

=

0 R

c.

Inoperative 1.(2 x 10 ) cps 5 (5 x 10 ) cps b.

Upscale NA NA

[

d.

Downscale 1 (3) cps 1 (2) cps e$

4.

INTERMEDIATE RANGE MONITORS i

a.

Detector not full in NA NA b.

Upscale 5 (108/125) divisions of 1 (110/125) divisions of full scale full scale c.

Inoperative NA NA d.

Downscale 1 (5/125) divisions of 1 (3/125) divisions of full scale full scale

'\\

5.

SCRAM DISCHARGE VOLUME a.

Water Level-High

$(

) inches 5(

) inches b.

Scram Trip Bypass NA NA 6.

REACTOR COOLANT SYSTEM RECIRCULATION FLOW a.

Upscale 1 (108/125) divisions of i (111/125) divisions of full scale full scale b.

Inoperative NA NA i

c.

(Comparator) (Downscale) 5 (10)% flow deviation 1 (11)% flow deviation 1

j j

• The Average Power Range Monitor rod block function is varied as a function of recirculation loop flow (W).

The trip setting of this function must be maintained in accordance with Specification 3.2.2.

=-

O TABLE 4.3.6-1

~

-Q m

CONTROL R00 BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS 9

g CilANNEL OPERATIONAL N

CilANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH O

TRIP FUNCTION CllECK TEST CALIBRATION (a)

SURVEILLANCE REQUIRED 1.

R0D BLOCK MONITOR a.

Upscale NA S/U(b)(c)

(c) q j.

b.

Inoperative NA S/U NA 1*

c.

Downscale NA S/U Q

18 2.

APRM

\\

d.

flow Blased Neutron Flux -

Upscale (NA)

S/U

,M (Q) 1 b.

Inoperative NA S/U(b),M NA 1, 2, 5 c.

Downscale (NA)

S/U H

IO)

I d.

Neutron Flux - Upscale, Startup (NA)

,S/U(b)'M (Q) 2, 5

{

3.

SOURCE RANGE MONITORS ye a.

Detector not full in NA S/U

,W NA 2, 5 m

b.

Upscale NA S/U(b),W Q

2, 5 c.

Inoperative NA S/U(b),W NA 2, 5 d.

Downscale NA S/U

,W Q

2, 5 4.

INTERMEDIATE RANGE MONITORS a.

Detector not full in NA S/U

,W NA 2, 5 b.

Upr,cale NA S/U(b),W Q

2, 5 c.

Inoperative NA S/Ug),W NA 2, 5 i

d.

Downscale NA S/U

,W Q

2, 5 5.

SCRAM DISCHARGE VOLUME a.

Water Level-Nigh NA (M) (Q)

R 1, 2, 5**

b.

Scram Trip Bypass NA M

NA (1, 2,) 5**

6.

REACTOR COOLANT SYSTEM RECIRCULATION FLOW a.

Upscale NA S/U

,M Q

1 b.

Inoperative i

NA S/U NA 1

S/U(b),M c.

(Comparator) (Downscale)

NA

,M Q

1 e

8

c..

. u TABLE 4.3.6-1 (Continued)

CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE RE0VIREMENTS NOTES:

a.

Neutron detectors may be excluded from CHANNEL CALIBRATION.

b.

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to startup, if not performed within the previous 7 days.

c.

Includes reactor manual control multiplexing system input.

With THERMAL POWER > (30)% of RATED THERMAL POWER.

With more than one control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

d*

O e

l l

l l

GE-STS (BWR/4) 3/4 3-52

l J

l REFUELfNG OPERATIONS BASES 3/4.9.6 REFUELIN_G PLATFORH The OPERABILITY requirements ensure that (1) the refueling platform will be used for handling control rods and fuel assemblies within t'he reactor pressure vessel, (2) each crane and hoist has sufficient load capacity for handling fuel assemblies and control rods, and (3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL The restriction on movement of loads in excess of the nominal weight of a fuel assembly over other fuel assemblies in the storage pool ensures that in the ' event this load is dropped 1) the activity release will be limited to that contained in a single fuel assembly, and 2) any possible distortion of fuel in the storage racks will not result in a critical array.

This assumption is consistent with the activity release assumed in the safety analyses.

l 3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL -SPENT FUEL STORAGE POOL The restrictions on minimum water level ensure that sufficient water depth is available to remove (99)% of the assun ed (10)% iodine gap activity released from the rupture of an irradiated fuel assembly.

This minimum water depth is consistent with the assumptions of the accident analysis.

3/4.9.10 CONTROL ROD REMOVAL These specifications ensure that maintenance'or repair of control rods or control rod drives will be performed under conditions that' limit the probability of inadvertent criticality.

The requirements for simultaneous removal of more than one control rod are more stringent since the SHUTDOWN MARGIN specification provides for the core to remain subcritical with only one control rod fully withdrawn.

3/4.9.11 RESIDUAL dEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal loop be OPERABLE or that an alternate method capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation ensures that 1)-suf-ficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140 F as required during REFUELING, and 2) sufficient coolant circulation would be available through the reactor core to assure accurate temperature indication and to distribute and prevent l

stratification of the poison in the event it becomes necessary to actuate the standby liquid control system.

The requirement to have two shutdown cooling mode loops OPERABLE when there is less than (23) feet of water above the reactor vessel flange ensures that a single failure of the operating loop will not result in a complete. loss of resid-ual heat removal capability.

With the reactor vessel head removed and (23) feet of water above the reactor vessel flange, a large heat sink is available for core cooling.

Thus, in the event a failure of tne operating RHR loop, adequate time is provided to initiate alternate methods capable of decay heat removal l

or emergency procedures to cool the core.

GE-STS (BWR/4)

S 3/4 9-2

'. 9 c

R'EFUELING OPERATIONS 3/4.9.10 CONTROL R00 REMOVAL SINGLE CONTROL ROD REMOVAL LIMITING CONDITION FOR OPERATION 3.9.10.1 One control rod and/or the associated control rod drive mechanism cay be removed from the core and/or reactor pressure vessel provided that at least the following requirements are satisfied until a control rod and associ-ated control rod drive mechanism are reinstalled and the control rod is fully inserted in the core.

The reactor mode switch is OPERABLE and locked in the Shutdown position a.

or in the Refuel position per Table 1.2 and Specification 3.9.1.

b.

The source range monitors (SRM) are OPERABLE per Specification 3.9.2.

The SHUT 00WN MARGIN requirements of Specification 3.1.1 are satisfied, c.

except that the control rod selected to be removed; 1.

May be assumed to be the highest worth control rod required to be assumed to be fully withdrawn by the SHUTDOWN MARGIN test, and a

2.

Need not be assumed to be immovable or untrippable.

d.

All other control rods in a five-by-five array centered on the control rod being removed are inserted and electrically or hydraulically disarmed or the four fuel assemblies surrounding the control rod or control rod drive mechanism to be removed from the core and/or reactor vessel are removed from the c~ ore cell.

e.

All other control rods are inserted.

APPLICABILITY:

CPERATIONAL CONDITIONS 4 and 5.

ACTION:

With the requirements of the above specification not satisfied, suspend removal of the control rod and/or associated control rod drive mechanism from the core and/or reactor pressure vessel and initiate action to satisfy the above requirements.

l l

l l

t i

GE-STS (BWR/4) 3/4 9-12

^

REFUELING OPERATIONS

+

l SURVEILLANCE REQUIREMENTS I

4.9.10.1 Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to the start of removal of a control rod and/or the associated control rod drive mechanism from the core and/or reactor pressure vessel and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereaf ter until a control rod and associ-ated control rod drive mechanism are reinstalled and the control rod is inserted in the core, verify that:

The reactor mode switch is OPERABLE and locked in the Shutdown position a.

l or in the Refuel position with the "one rod out" Refuel position interlock OPERABLE per Specification 3.9.1.

A b.

The SRM channels are OPERABLE per Specification 3.9.2.

l The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied c.

per Specification 3.9.10.1.c.

l d.-

All other control rods in a five-by-five array centered on the control rod being removed are inserted and electrically or hydraulically disarmed or the four fuel assemblies surrounding the control rod or i

control rod drive mechanism to be removed from the core and/or i

reacter vessel are removed from the core cell.

e.

All other control rods are inserted.

I 4

i I

i l

l l

GE-STS (BWR/4) 3/4 9-13

~

-_m 4

m _ _

..._._..__.m._-

_.m--

.u-4.-

..1.

' d j

REpVELINGOPERATIONS MULTIPLE CONTROL ROD REMOVAL LIMITINGCONDITIONFOROPERATION 3.9.10.2 Any number of control rods and/or control' rod drive mechanisms may be removed from the core and/or reactor pressure vessel provided that at least the following requirements are satisfied until all control rods and control 4

rod drive mechanisms are reinstalled and all control rods are inserted in 1

core.

The reactor mode switch is OPERABLE anc locked in the Shutdown position n.

or in the Refuel Refuel position " position per Specification 3.9.1, except that the for those control rods and/or control rod drive mechanisms to b i

removed below,, after the fuel assemblies have been removed as specified b.

The source range monitors (SRM) are OPERABLE per Specification 3.9.2.

The SHUTDOWN-MARGIN requirements of Specification 3.1.1 are satisfied.

c.

d.

All other control rods are either inserted or have the surrounding four fuel assemblies removed from the core cell.

2 The four fuel assemblies. surrounding each control rod or control rod e.

drive mechanism to be removed from the core and/or reactor vessel are removed from the core cell.

[

APPLICABILITY:

OPERATIONAL CONDITION 5.

4 ACTION:

With the requirements of the above specification not satisfied, suspend removal of control rods ar.d/or control rod drive mechanisms from the core and/or pressure vessel and initiate action to satisfy the above requirements.~~ -

l GE-STS (BWR/4) 3/4 9-14 f- - -.

~,. ;

s.c u

~

?-

~

REFUELING OPERATIONS-

~

s SURVEILLANCE RE0VIREMENTS J

s 4

4.9.10.2.1 Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior 'to the startN[f removal of control rods and/or control rod drive mechanisms from'the.gors and/or reactor pressure vessel and at least once per 24 drive mechanisms are. hours thereafter'until all' control rods and contro reinstalled and al;l control rods are inserted in the core, verify that:

s The reactor Gode switch is OPE'RABLE and locked in the Shutdown positiqn I a.

or in the Refuel position per Specification 3.9.1.

sl s

b.

The SRM channels are OPERABLE per Specification 3<9.2.

c.

The SHUT 00KN' MARGIN requirements of Specification 3.1.1 are satisfied.

d.

All other control rods are either inserted or have the surrounding four fuel assemblies removed from the core cell.

The four fuel assemblies surrounding each control rod and/or control e.

rod drive mechanism to be removed from the core and/or reactor yessel are removed from.the core cell.

4.9.10.2.2 Following replacement of all control rods and/or control rod drive mechanisms removed in accordance with this specification, perform'a. functional ~

test of the "one rod-out" Refuel position interlock, if thir function had been bypassed.

GE-STS (BWR/4) 3/4 9-15

\\

9

2 INSTRUMENTATION l

BASES 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentation is provided to initiate actions to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to i

the reactor vessel without providing actuation of any of the emergency core cooling equipment.

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 less than the drift allowance assumed for each trip in the safety analyses.

3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION l

The control rod block functions are provided consistent with the require-ments of the specifications in Section 3/4.1.4, Control Rod Program Controls and Section 3/4.2 Power Distribution Limits.

The trip logic is arranged so that a trip in any one' of the inputs will result in a control rod block.

Operation with a tr.ip set less conservative than its Trip setpoint but within its specified Allowable Value is acceptable on the basis that the t

difference 'between each Trip Setpoint and the Allowable Value is equal to or i

i less than the drift allowance assumed for each trip in the safety analyses.

f 3/4.3.7 MONITORING INSTRUMENTATION f

3/4.3.7.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring instrumentation ensures that; j

(1) the radiation levels are continually measured in the areas served by the i

individual channels; (2) the alarm or automatic action is initiated when the radi-ation level trip setpoint is exceeded; and (3) sufficient information is avail-able on selected plant parameters to.tonitor and assess those variables follow-ing an accident.

This capability is consistent with the recommendations of (NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980).

3.4.3.7.2 SEISMIC MONITORING INSTRUMENTATION The OPERABILITY of the seismic monitoring' instrumentation ensures that suf-ficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety.

This j

capability is required to permit comparison of the measured response to that l

used in the design basis for the unit.

(This instrumentation is consistent with the recommendations of Regulatory Guide 1.12 " Instrumentation for Earthquakes",

April 1974.)

3/4.3.7.3 METEOROLOGICAL MONITORING INSTRUMENTATION The OPERABILITY of the meteorological monitoring instrumentation ensures that sufficient meteorological data is available for estimating potential radia-tion doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere.

This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public.

(This instrumentation is consistent with the recommenda-tions of Regulatory. Guide 1.23 "Onsite Meteorological Programs," February 1972.)

j GE-STS (BWR/4)

B 3/4 3-4 l

j a

3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:

Preserve the integrity of the fuel cladding.

a.

b.

Preserve the integrity of the reactor coolant system.

Minimize the energy which must be adsorbed following a loss-of-coolant c.

accident, and d.

Prevent inadvertent criticality.

This specification provides the limiting conditions for operation necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be out of service because of main-tenance.

When necessary, one channel may be made inoperable for brief intervals to conduct required surveillance.

The reactor protection system is made up of'two independent trip systems.

There are usually four channels to monitor each parameter with two channels in each trip system.

The outputs of the channels in a trip system are combined in a logic so that either channel wilil trip that trip system. The tripping of both trip systems will produce a reactor scram.

The system meets the intent of IEEE-279 for nuclear power plant protection. systems.

The bases for the trip settings of the RPS are discussed in the cases for Specification 2.2.1.

The measurement of response time at the specified frequencies provides assurance that the protective functions associated with each channel are com-pleted within the time limit assumed in the accident analysis.

No credit-was taken for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential, overlap'p'ing or total channel test measurement, provided such tests demonstrate the total channel response time as defined.

Sensor response time verification may be demonstrated by either (1) inplace, onsite or offsite test measurements, or (2) utilizing replacement sensors with certified response times..

GE-STS (BWR/4)

B 3/4 3-1

-,e