ML20006C032

Amend 77 to License DPR-34,upgrading Portions of Tech Specs Re Reactivity Control,Including Calculated Bulk Core Temp, Core Average Inlet Temp,Core Alterations & Control Rod Pair Operability

ML20006C032
Person / Time
Site:	Fort Saint Vrain
Issue date:	01/24/1990
From:	Weiss S Office of Nuclear Reactor Regulation
To:
Shared Package
ML20006C016	List: ML20006C015 ML20006C032 ML20006C035
References
NUDOCS 9002060218
Download: ML20006C032 (75)

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UNITED STATES NUCLEAR REGULATORY COMMISSION o

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.l wAsHmotow.o.c.2oses PUBLIC SERVICE COMPANY OF COLORADO DOCKET NO. 50-267 FORT ST. VRAIN NUCLEAR GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 77 License No. DPR-34 1.

The Nuclear Regulatory Comiision (the Comission) has found that:

A.

The application for amendment by Public Service Company of Colorado (thelicensee)datedSeptember 14, 1989 as revised October 13, October 30 and December 4, 1989, complies with the standards and requirementsoftheAtomicEnergyActof1954,asamended(theAct),

and the Commission's regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, as amended.the provisions of the Act, and the regulations of the Comission; C.

There is reasonable assurance:

(1)thattheactivitiesauthorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

9002060218 900124 PDR ADOCK 05000267 P

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2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachw nt to this license amendment, and paragraph 2.D.(2) of Iacility Operating License No. DPR-34 is hereby amended to read as follows' (2)' Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 77, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

t 3.

The license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY' COMMISSION

.. r m Seymour 11. Weiss, Director Non-Power Reactor Decommissioning and Environmental Project Directorate Division of Reactor. Projects - III, IV, V and Special Projects Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications-Date of Issuance:

January 24, 1990 L

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l ATTACHMENT TO LICENSE AMENDMENT NO. 77 TO FACILITY OPERATING LICENSE.NO. DPR 34 DOCKET NO. 50-267 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by amendment number and i

contain vertical lines indicating the areas of change.

Remove Insert 2-2 2-2 2-5 2-5 2-9 2-9

.2-10 4.1-2(markout) 4.1-2 4.1-3 through 4.1-4 4.1-11 4.1-5 4.1-6 4.1-7 4.1-8 4.1-9 4.1-10 4.1-11 5.1-1 5.1 5.1-2 through 5.1-3 5.1-7 5.1-4 5.1-5 5.1-6 5.1-7 5.1-8 5.1-8 5.1-9 5.1-9 5.1-10 5.1-10 5.1-11 5.1-11 5.1-12 0.1-12 5.1-13 S.1-13 5.1-14 5.1-14 l

5.1-15 5.1-15 l

5.1-16 5.1-16 l

.7.1-16 7.1-16 Reactivity Control Section l

l 1-1 through 1-2 l

3/4-0-1 through 3/4 0'-11 3/4 1-1 through 3/4 1-52 l'

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3 F rt St. Vrain 01 Technical Specifications Amendment No. 77 Page 2-2 2.2 Eauipment Surveillance Test

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J A test of the functional capability cf a piece of equipment to determine that it is operable, This may consist of either an on line or off line demonstration of i

the operability of the equipment.

2.3 Instrumentatien Surveillance a) Channel Check r'

A qualitative determination that the channel -is operable. The determination is made by observation of channel behavior during operation or comparison with other channels monitoring the same variable or related variables.

l b) ChannelTest(CHANNELFUNCTIONALTEST) l A test of the functional capability of the channel to determine that it is operable. This may consist of the injection of a simulated signal into a channel as close as possi.ble to the primary sensor to verify that it is operable.

c) Channel Calibration The adjustment of a channel so that it corresponds within acceptable range and accuracy, to known values of the parameter which the channel monitors.

Calibration shall encompass the channel and alarms up to the bistable output.

2.4 Irradiated Fuel Irradiated fuel is fuel that has a radiation level 2 100 mr/hr measured one foot from the element surface.

2.5 Low power Operation Low Power Operation is any operation with the, Wide Range Logarithmetic instrumentation indicating greater than l

l 10E-3 and less than 2% of rated thermal power.

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~ Technical. Specifications' Amendbent No. 77 1

E Page 2-5

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'2 16. Refueling Shutdown'

'1 The reactor is considered shut down for refueling purposes.

l when the reactor mode switch is locked in the " Fuel Leading" pcsition simultaneous with either hot shutdown,or j

the cold shutdown reactivity conditions.

2.17 Safy"ShutdownCooling Safe shutdown cooling refers to cooling of the cor6 with J

Safe Shutdown Equipment providing for removal of core stored energy and for adequate sustained decay heat removal. The reactivity condition in the core is either hot or cold shutdown.

2.18 Surveillance Interval A surveillance interval is the interval of time between rurveillance check, tests, or calibration.

Unless ctherwi se

stated, the surveillance interval can be adjusted by 25% to accommodate normal operational schedules.

Unless otherwise stated in these specifications, surveillance may be terminated on those instruments or equipment not in normal use during reactor l

shutdown or refueling shutdown.

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-Technical Specif.1 cations c

Amendment No. 77 Page 2-9 2.23.-CALCULATED BULK CORE TEMPERATURE The CALCULATED BULK CORE TEMPERATURE (CBCT) shall be the-calculated average temperature of the core, incl udi ng --

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graphite and fuel', but not the reflector, assuming a -loss I

of all forced circulation of PRIMARY COOLANT-FLOW. Use of

-l the CA1.CULATED ~ BULK CORE TEMPERATURE is explained in' l

LCO 4.0.4.

s, 2.24 CORE AVERAGE INLET TEMPER'ATURE The CORE AVERAGE INLET TEMPERATURE shall be the arithmetic average of the operating circulator inlet temperatures,-

adjusted 'for circulator power input, steam generator l-regenerative neat input, and heat transfer to. (or from)

I the PCRV liner cooling system.

l 2.25' ACTION l

ACTION shall be that part of a specification which l

prescribes remedial measures required under designated ~

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conditions.

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2.26 CORE ALTERATION (S) l CORE ALTERATION (S) shall be the movement or manipulation l

of any component within the PCRV that alters the core j

l reactivity (except for insertion of control rod pairs or l

reserve. shutdown material) or activities that could result l

in damage to the core components, while fuel is in the l

reactor vessel.

Suspension of CORE ALTERATION (S): shall I

l not-preclude completion of movement of a component to a

-l safe. conservative position or condition.

l 2.27 CORE AVERAGE TEMPERATURE l

a.

During SHUTDOWN and REFUELING, CORE AVERAGE f

l TEMPERATURE shall be the arithmetic average of the l

CORE AVERAGE INLET TEMPERATURE and the CORE AVERAGE l

OUTLET TEMPERATURE.

I b.

During STARTUP, LOW POWER, and POWER, CORE AVERAGE I

I TEMPERATURE shall be thermodynamically calculated 1

-l based on CORE AVERAGE INLET and CORE AVERAGE OUTLET l

TEMPERATURES, PRIMARY COOLANT FLOW, and THERMAL POWER.

l 2.28 POWER-TO-FLOW RATIO (P/F) l POWER-TO-FLOW RATIO (P/F) shall be the percentage of RATED l

THERMAL POWER divided by the percentage of design PRIMARY l

COOLANT FLOW at RATED THERMAL POWER.

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Feet St. Vrain #1'

-Technical Specifications.

Amendment No. 77 Page 2-10 l

2.29 - [RIMARY COOLANT FLOW l

The PRIhAF," COOLANT FLOW shall be the sum of the helium l

massflow (!b/hr) for each of the OPERATING circulators, l

The design PRIMARY COOLANT FLOW at RATED THERMAL POWER is l.

3.5E+06 Lib /hr.

l 2.30 SHUTOOWN MARGIN l

SHUTOOWN~ MARGIN 11 be the instantaneous amount'of l

reactivity by which the reactor is subcritical or would be l

suberitical from its present condition assuming.that all l

OPERABLE control rod pairs are fully inserted except for l

the single control rod pair of highest reactivity worth I

capable of being withdrawn, which is assumed to be fully l

withdrawn.

l 2.31 THERMAL POWER l

THERMAL POWER _ shall be the total reactor core heat l

transferred to the. reactor coolant, as determined by an

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appropriate heat balance calculation, or from calibrated l

nuclear instrumentation.

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. Technical Specifications e

4 Amencment No. 77_

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Page 4.1-2 through N-Page 4.1-11 REACTOR' CORE -AND REACTIVITY CONTROL LIMITINO CONDITIONS FOR OPERATION (Continued) l The following Specifications have been superseded by Specifications

- l in the Reactivity Control Section:

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LCO 4cl.4 l

LCO 4.1.5' l

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Technical Specifications.

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Amendment No. 77 i

Page 5.'l-1.through

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. 5,1' REACTOR CORE AND REACTIVITY CONTROL - SURVEILLANCE REQUIREMENTS _

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l The following Surveillance Requirements have been superseded by 4

-l Specifications in the Reactivity Control Section:

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SR 5.1.1

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l SR 5.1.2 I

SR 5.1.3 s-i l

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Technical: Specifications

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Amendment No. 77

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Page 5.1-3 i

Specification SR 5.1.4 - Reactivity Status Surveillance-t h

A surveillance check of the reactivity status of the core-

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shal'1 be performed at' each startup and once per week during power operation.

If the difference between the

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observed and the expected reactivity, based on normalization to. a base steady state core condition, reaches 0.01 am, this discrepancy shall be considered an abnormal occurrence.

The initial base steady state core condition and changes of this base shall bc approved by the NFSC.

Basis for Soecification SR S.I.4 The specified frecuency of the surveillance check of the core reactivity status will assure that the difference i

between. the observed and expected core reactivity will be i

evaluated regularly.

This specification is designed to ensure that the core reactivity level is monitored to reveal in a timely manner the existence of potential safety problems or operational l

problems. An unexpected and/or unexplained change in the t

observed ae coactivity could be. indicative of such problems.

The normali:st :n to an initial base steady state core i

condition

- '1 eliminate discrepancies due to manufacturing tolerances, analytict1 modeling w-

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Tecnnical Specifications:

, 41 E

Amendment No. 77.

.Page 5.1 :

approximations and deficiencies in basic data at the beginning of operation.

Changes of the base steady state 1

4 a-core conditions are permissible to eliminate explainable.

discrepancies resulting from long-term reactivity burnup' effects'and core refuelings.

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Comparison of predicted and observed reactivities in a

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i base steady state configuration will ensure the comparison

- will be easily understood and readily evaluated.

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Any reactivity anomaly greater than 0.01 as would be

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unexpected and its occurrence would be thoroughly investigated and evaluated.

.The value of 0.01 An is considered to be a safe limit since a shutdown margin of at least 0.01'Ac with the highest worth rod pair fully l

withdrawn is always maintained (see LCO 3.1'.4).

Specification SR 5.1.5 - Withdrawn Rod Reactivity Surveillance i

l This Surveillance Requirement has been superseded by l

Specifications in the Reactivity Control Section.

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Technica1'3pecifications Amendment No.?77 Page 5.1-10 1

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Specification SR 5.1.6 - Core Safety Limit surveillance-During power operation 'the total operating time of_the fuel elements within the core at power-to-flow ratios u,

above the curve of Figure 3.1-2 will be evaluated once per L<

week when the plant operation is within the-normal i

operating-range, and as soon as practicable after any.

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- deviation. from the normal- _ operating range.

These operating times' will be compared to. the allowable' operating time of Specification SL 3.'1 to assure that-the Core Safety Limit has not been exceeded.

j Basis for Specification SR 5.1.6 Only during operation of the plant outside of the normal operating range is there a potential for accumulating-significant operating times at' power-to-flow ratios greater than the curve of Figure 3.1-2.

Therefore, weekly evaluations of the total accumulated operating time at power-to-flow ratios greater than the curve of Figure 3.1-2 is sufficient during normal operation.

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Technical Specificasions.

'I Amendment No. 77-Page S.1 I Following any significant. deviation from the' normal operating ~ range, the operation should be evaluated to

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' determine the-degree _to which the actual total operation I

of the core approached the Core Safety Limit.

Specification SR 5.1.7 - Recion Peaking Factor i

Surveillance i

s The. calculated region peaking factors (RPF's) used in determining the individual region outlet temperatures for Regions 20 and 32 through.37 and percent RPF. discrepancy.

(see LCO 4.1.7) for Regions a through 19 and 21 through 31-I shall be evaluated according to the following schedule for

.i each refueling cycle:

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Calculated RPF's:

1)

Prior to initial power D

operation after

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refueling.

2)

At the equivalent of 20 (2 5) effective days at.

rated thermal power after refueling.

3)

At the equivalent of 40

(*})effectivedays at rated thermal power after refueling.

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,eennical-. Specifications.

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Amendment No 77 Page 5.1-12 4)

At monthly intervals thereafter.

'provided

.that-the-core.

has accumula'ted an exposure-of at least-the equivaler.t of.

10 effective days..at

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. rated thermal power.

s since the previous evaluation.

If-the core

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has accumulated-an.

exposure of less than

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the equivalent o f.

1 10 effective days at r

rated thermal power since the previous evaluation, the evaluation' may.

be deferred until the next applicable interval, b)

Percent RPF Discrepancy:

Within a total elapsed time of 10 calendar days at. reactor power levels above 40%

of rated thermal power after the completion of any of the l

' Fort St.'Vrain #1"

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-Technical Specifications-

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' Amendment No. 77-~

Page 5.1-13 '

" Calculated-

.RPF" i

evaluations required above with the fcilowing qualifications:

1)

A

" Percent RPF Discrepancy" evaluation shall be performed prior to exceeding 40%

of rated thermal power for 4

the first time after refueling, bu t' - at a l-reactor power above 30%

of rated thermal power.

2)'

If

,the total elapsed time at reactor power levels above 40% of.

rated thermal power does not-exceed 10 calendar days prior to the subsequent

" Calculated RPF" evaluation, the

"- Percent RPF.

Discrepancy" evaluation is not required, but the total elapsed time at reactor power levels 4

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Op r;'t at. eretn~as Technical Specifications:-

j' Amendment No. 77s y

Page:5.1,

above~

401 of' rated I

i thermal power between.

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" Percent-RPF

' Discrepancy" svaluations shall noti exceed 45 calendar days.

Basis for Specification SR 5.1.7.

The calculated region peaking factors for Regions 20 and--

32 through 37 and their comparison regions will-change during the refueling cycle as fission product-inventories 1

saturate, fissile material and burnable poison are depleted, and control rods are withdrawn from the core.

1 Evaluations based upon operating experience gained prior to completion of rise-to power testing (i.e.,- Cycles 1 and.

2 and part of Cycle 3) indicate that the ratio of the calculated region peaking factors -in Regions 20 and 32 through 37 to the calculated region peaking. factors in comparison regions as a

function of control rod configuration, changes gradually in a predictable manner during a refueling cycle.

A surveillance check of the calculated region peaking factors at the specified frequency will assure that the appropriate region peaking

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factors continue to be used in detfrmining the region outlet temperature for Regions 20 and 32 through 37.

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Technical Specifications

. Amendment No. ' 77

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Page 5.1-152 i

40 The calculated and measured. region peaking f.cters for.

Regions 1. through -19 and 21 through 31' (candidate comparison regions) will change during the refueling cycle I

1 as fission product inventories saturate,; fissile material ~

i and burnable -poison are depleted, control rods are withdrawn from the core, and. region flow-characteristics

-change.

A surveillance check: of the percent region I

peaking. factor discrepancy will provide assurance that the requirements of LCO 4.1.7c are being met for comparison regions.

The frequency for surveillance has been established based upon conservative evaluations of potential fuel kernel migration, which could occur.if-a region with an excessively large, negative region peaking factc~ discrepancy were used as a comparison region.

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g,J Amer.dment No.' 77 Page 5.1-16 P

SPECIFICATION SR 5.1.B MINIMUM HELIUM FLOW / MAXIMUM CORE REGION-TEMPERATURE RISE SURVEILLANCE RE0u!REMENT

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The total helium circulater flow or the helium coolant temperature rise through each core region shall be determined to be within the limits of LCO 4.1.9 at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

BASIS for SPECIFICATION SR 5.1.8 Surveillance of the-helium circulator flow or ~ helium' coolant temperature rise once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ensures that the reouirements of i

LCO 4.1.9 are met.

In addition, plant procedures require that the flow rate, core outlet temperatures, and power level be monitored continuously whenever the power level is being changed or orifice b

valves are being adjusted.

In performance of the surveillance, the

. total reactor helium coolant flow is determined by calculation i

consistent with the method used to determine the required flow for L

the analysis l-l

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v Technical Specifications

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Amendment No. 77-Page 7.1-16 (7) All Reportable Events.

(8)'Any indication that there may be a deficiency in-some Easpect.

of design or operation of structures,- systems, or components, that affect nuclear safety, t

(9)' Reports and' meeting minutes of the PORC.

b.

.The Nuclear Facility Safety. Committee shall approve:

(1) The control rod withdrawal sequence as required

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-l by Technical Specification 3/4.1.5.

(2) The initial base (reactivity) steady state core -

condition and changes to the base (reactivity) as required by Technical Speci fi car. ion LCO 4.1.8.

-NFSC permission is required before reactor operations may. resume if a reactivity anomaly of 0.01 Delta K is reached.

(3) Proposed.

changes to

facility, operating procedures and tests or-experiments that are determined to involve an unreviewed

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environmental question.

c.

Audits of facility activities shall be performed under the cognizance of the Nuclear Facility Safety Committee. These audits shall encompass:

(1) The conformance of facility ~ operation to all provisions contained within the

-Technical Specifications and applicable license conditions at least once per year.

i (2) The performance, training, and qualifications, of the facility staff at least once per year.

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(3) The results of actions taken to-correct deficiencies occurring in. facility equipment, structures, systems, or method of operation that affect nuclear safety at least once per six months.

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REACTIVITY CONTROL SECTION 4

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-Fort St. Vrain cl-1-

'e Technical Specifications Amendment No, 77

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-Page 1-1

<e DEFINITIONS

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j 1.0 DEFINITIONS-L l The definitions of OPERATIONAL MODES shown in Table 1.0-1 for the

6' l Reactivity Control Technical Specifications are different from the l definitions of various power levels for the Non-Reactivity Control

l. Section Technical Specifications, t

'/I l OPERATIONAL MODE-MODE l 1.1 An OPERATIONAL MODE (i.e.

MODE) shall correspond to any one l

inclusive combination of Reactor Mode Switch Setting, Inter!ock t

l Sequence Switch Setting, and f4 RATED THERMAL POWER, specified i

in Table 1.0-1~.

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Technical Specifications Amendment. No. 77 -

Page 1-2 b

OEFINITIONS-l TABLE 1.0-1 l

OPERA 1IONAL MODES N

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4 l-INTERLOCK

. REACTOR l

SE0VENCE

' MODE SWITCH

% RATED l-MODE SWITCH SETTING SETTING

-THERMAL POWER

• 1 I

l POWER Power Run

> 30%

l OPERATION (P).

l LOW POWER (L)

Low Power 0 Run

> 5% and 5 30%-

l STARTUP (S/U)

Startup Run 5 5%

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SHUTOOWN (S/0)

Off #

0 I

REFUELING (R)

Fuel Loading 0

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Excluding decay he't.

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Interlock Sequence Switch (ISS) may be in any position in l

SHUTOOWN and REFUELING.

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Includes Reactor Internal Maintenance.

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1. The Reactor Mode Switch setting may be changed for the purpose of l

performing surveillances or other tests, provided the control rods t

l are verified to remain fully inserted (or as otherwise required l

for Refueling operations or surveillance testing) by a second l

licensed operator or other qualified member of the unit technical l

staff.

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@ The Interlock Sequence Switch setting may be changed to the POWER l

position and power may be increased to no greater than 40%, for l

the purpose of performing surveillances or ether tests, for up to l

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, without being considered a change in OPERATIONAL MODES.

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Amendment No. 77 Page 3/4 0-l'

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l 3/4.0 REACTIVITY APPLICABILITY l These-Applicability Specifications are specifically applicable to

-l Reactivity Control Specifications 3/4.1.1 through 3/4.1.9. and shall l not be used for any Non-Reactivity Control Section Specifications.

l 3.0 REACTIVITY LIMITING CONDITIONS FOR OPERATION l 3.0.1 Compliance with the Limiting Conditions for Operation l

contained in the succeeding specifications is required I

during the OPERATIONAL MODES or other conditions specified l

therein, except that upon failure. to meet the Limiting l

Conditions for. Operation, the associated ACTION requirements j

shall be met.

l 3.0.2 Noncompliance with a specification shall exist when the I

requirements of the Limiting Condition for Operation and i

I associated ACTION requirements are not met within the i

l specified time intervals.

If-the Limiting Condition for l

Operation is restored prior to expiration of the specified l

time interval, completion of the ACTION requirements is not

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required.

-l l 3.0.3-When a Limiting Condition for Operation is not. met, except l

as provided in the associated ACTION requirements. ACTION

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shall be initiated within I hour to place the unit in an i

l OPERATIONAL MODE in which the specification'does not apply l

l by placing it, as applicable', in at least LOW POWER within-l the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in at least SHUTDOWN within the l

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

This requirement is not applicable in l

SHUTDOWN or REFUELING.

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l 3.0.4 Entry into an OPERATIONAL MODE or other specified condition l

l shall not be made when the conditions for~ the Limiting j

l Condition for Operation are not met and the' associated l

ACTION requires a shutdown if they are not met within a l

specified time interval.

Entry into an OPERATIONAL MODE or l

specified condition may be made in accordance with ACTION l

requirements when conformance to them permits continuea l

operation of the facility for an unlimited period of time.

l This provision : hall not prevent passage through or to l

OPERATIONAL MODES as required to comply with ACTION I

requirements, or as required by automatic or manual l

protective ACTION.

Exceptions to these requirements are l

stated in the individual specifications.

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-Technical Specifications Amendment No' 77-

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Page 3/4 0-2 1.

-l 3.0 LIMITING CONDITIONS FOR OPERATION (Continued) l'3.0.5 Specifications identifying requirements in terms of j

l CALCULATED BULK CORE TEMPERATURE (CBCT) shall be in j

1-accordance with the discussion in LCO-4.0.4.

j l 4.0 REACTIVITY SURVEILLANCE REQUIREMENTS q

I l 4.0.1 Surveillance Requirements shall be applicable only during l

l the OPERATIONAL MODES or other conditions specified for l

. individual Limiting Conditions for Operation unless 3

l otherwise stated in an individual Surveillance Requirement.

l 4.0.2 Each Surveillance Requirement shall be performed within the l

specified time ' interval per Definition 2.I8,

" SURVEILLANCE l

INTERVAL".

l 4.0.3 Failure to perform a Surveillance Requirement within the I

allowed SURVEILLANCE INTERVAL, defined by SR 4.0.2, shall l

constitute noncompliance with the OPERABILITY requirements i

for a Limiting Condition for Operation.

The time limits of l

the ACTION requirements are applicable at the time it is l

identified that a-Surveillance Requirement has not been l

. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to permit the completion of the surveillance when performed.- The ACTION requirements may be delayed for up to l

l the allowable outage time limits of the ACTION requirements j

l are less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Surveillance Requirements do not i

l have to be performed on inoperable equipment, i

y l 4.0.4 Entry into an OPERATIONAL MODE or other-specified condition l

shall not be made unless the Surveillance Requirement (s) l associated with the Limiti'ng Condition for Operation have l

been performed -within the stated SURVEILLANCE INTERVAL.

j l

This provision shall not prevent' passage through or to j

l OPERATIONAL MODES as required to comply with ACTION l

requirements.

Exceptions to these requirements are stated l

in the individual specifications.

Fort St. Vraib dl; e'

N Technical Specifications-Amendment No. 77' L'

Page 3/4 0-3 i

t

~

l-BASIS FOR SPECIFICATION LCO 3.0/SR 4.0_

i l The specifications of this section provide the general requirements l applicable to each of the Limiting Conditions for Operation and

.l. Surveillance Requirements within the' Reactivity Control Section.'

l-These - requirements are based on the requirements for Limiting

'l Conditions -for Operation stated in the Code of Federal Regulations, l 10 CFR 50.36(c)(2):

l " Limiting conditions for_ operation are the lowest functional l capability or performance levels of equipment. required-for safe I operation of the facility. When a limiting condition for operation l of'a nuclear reactor is not met, the licensee shall shut d o.wn. the l reactor or follow -any remedial _ action permitted by the technical l specification until the condition can be met."

l These Limiting Conditions for Operation provide for operation with l sufficient redundancy and/or diversity to meet the single-failure I cr_iterion as relied upon in the plant's safety analysis. -The l Limiting Conditions for Operation do-not replace plant operating i procedures.

Plant operating procedures establish plant operating l conditions with at least the capability and performance specified in

.I these Limiting Conditions for Operation.

l 3.0.1 This specification defines the applicability _of each I

specification in terms of defined OPERATIONAL MODES or other l

specified conditions and is provided to delineate l

specifically when each specification is applicable.

l The ACTION requirements establish those remedial measures k

l that must be taken within specified time-limits when the l

requirements of a Limiting Condition for Operation are not

'l met.

4

V FertL5t. V* aim #1

~

li( g3 ; 6 -

Technical Specifications j 1" "

Amendment No. 77:

'O *M Page 3/4 0 ;

l' BASIS FOR SPECIFICATION LCO 3' 0/SR 4.0 (Continued) l There are two basic types of ACTION requirements. The first I

specifies 'the remedial measures that permit continued l

operation of the facility which is not further restricted by l

l the time limits of the ACTION requirements. An example of 1

l this is the ACTION to.be taken for inoperable'seismicL l

l monitors.

In this'. case, conformance. 'to the ACTION l

requirements-provides an acceptable level of safety for l

unlimited continued operation as long as the ACTION i

l requirements continue to be met.-

The second type of ACTION l

requirement specifies a time limit in which conformance to l

the conditions of the Limiting Condition for Operation must j'

l-be met.

This time limit is the allowable outage time to l

restore an inoperable system or component to OPERABLE status l

or for restoring parameters within specified l i mi t s ~.

If l

these actions are not completed within the allowable. outage l

time limits, a shutdown is required to place the facility in I

a MODE or condition in.which the specification no longer l

applies.

It is not intended that the shutdown ACTION l

requirements be~ used as an operational convenience which-1 permits (routine) voluntary removal of a system (s)- or l

l component (s) from service in lieu of other alternatives that l

would not result in redundant systems or components being l

inoperable.

l The specified time limits of the ACTION requirements are l

applicable from'the point in time it is identified that a l

Limiting Condition for Operation is ~not met..The time 1

l limits of the-ACTION requirements are also applicable when a l

system or component is removed from service for surveillance I

testing, including investigation, maintenance, repairs, nor l

modifications to resolve operational problems.

Individual l

specifications may include a specified time limit for the l

completion of a Surveillance Requirement when equipment is

-l removed from service.

In this case, the allowable outage l

time limits of the ACTION requirements are applicable when l

this limit ' expires if the surveillance has not been l

completed.

When a shutdown is required to comply with l

ACTION requirements, the plant may have entered a M005 in I

which a new specification becomes applicable. -In this case.

l the time limits of the ACTION requirements would apply from l

the point in time that the new specification becomes l

applicable if the requirements of the Limiting Condition for i

Operation are not met.

[

c..St.VraEnal i

N w

Technical Specifications Amendment No. 77 Page 3/4 0 <

n

l. BASIS FOR SPECIFICATION,LCO 3.0/SR 4.0 (Continued) i

?

l 3.0.2 This specification establishes that noncompliance with a l

specification exists when the requirements of the Limiting l

l Condition for Operation are. not met'- and the associated l

ACTION requirements have not been implemented. within the l-specified. time interval. ; The purpose of this specification l

1s to. clarify that (1) implementation of the ACTION j

l requirements within the specified time interval constitutes

~:

l compliance'with a specification and (2) completion ~ of the l

remedial measures of the ACTION requirements is not required I

when compliance with a Limiting Condition o f.

Operation is l

restored within the

. time i.n terva l specified in the l

associated ACTION requirements.

l This concept also applies to progressive ACTIONS.

For i

example, if an ACTION allows 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to repair one SLRDIS l

valve and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to repair all but one SLRDIS valve. (in l

the event several valves are inoperable), once the equipment l

1s restored to only one inoperable valve the original 72 l

hour. clock is applicable.

l 3.0.3 This specification delineates the measures to be taken fur i

those circumstances not directly provided for_ in the ACTION l

statements and whose occurrence would violate the-intent of I

a specification.

l The purpose of this specification is to delineate the time i

limits for placing the unit in a SHUTOOWN MODE when plant l

operation cannot be maintained within the limits for safe l

operation defined by the Limiting Conditions--for Operation l

and its ACTION reouirements.

It is not intended to be used 1

as an operational convenience which permits (routine)^

l voluntary removal of redundant systems or components from I

service in lieu of other alternatives that would not result-

-l in redundant systems or components being inoperable.

One I

hour is allowed to prepare for an orderly shutdown before-l initiating a change in plant operation.

This time permits-l the operator to coordinate the reduction in electrical l

generation with the load dispatcher to ensure the stability l

and availability of the electrical grid.

The time limits I

specified to reach lower MODES of operation permit the j

shutdown to proceed in a controlled and orderly manner that l

is well within the specified maximum cooldown rate and l

within the cooldown capabilities of the facility assuming l

cnly the minimum required equipment is OPERABLE.

This l

reduces thermal stresses on components and the potential for l

a plant upset that could challenge safety systems under l

conditions for which this specification applies.

-w-

~

Fort'St. Vraia dl

~ Technical Specifications

. Amendment No. 77

.Page 3/4 0-6 l BASIS FOR SPECIFICATION LCO 3.0/SR' 4.0 (Continued) l If remedial measures permitting limited continued operation l

of.the facility' under' the provisions of the ACTION l-requirements are completed, the shutdown may be terminated.

l

-The time limits of the ACTION requirements are applicable'

,?

l

- from the point in time. there was a failure to meet a l

Limiting Condition for Operation.

Therefore,- the shutdown l

may be terminated if the ACTION requirements have been met a

H or the time limits of the ACTION requirements have' not

expired, thus providing an allowance for-the completion of the required actions.

l The time limits of LCO 3.0.3 allow 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the plant to l.

be in SHUTDOWN when a shutdown 1s-required during plant l

operation.

However,.i f a lower MODE of operation-is l

reached in less time than allowed, the total allowable time l-to. reach SHUTDOWN,. or other applicable MODE, is not reduced.

l For example, if LOW POWER is reached in.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the time l

allowed. to reach SHUTDOWN is the next 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> because the l

total time to reach SHUTOOWN is not reduced from _the l

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

Therefore, if remedial

- l measures are completed that would permit a return to POWER l

operation-a penalty is not incurred for having reached a l

lower MODE of operation in less than the total time allowed.

l The ACTION to be in LOW POWER in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in SHUTDOWN in l

-the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> defines an orderly shutdown at Fort l

St. Vrain.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to reduce to LOW POWER (30%) is allowed l

to minimize unnecessary transients on the steam generator l

tubing that would result from going through boilout l

(approximately 18-22%) during reductions to lower. power l

levels.

The process of reducing power in an orderly manner i

from less than 30% (LOW POWER) to SHUTDOWN is complicated l'

and time consuming in that all of the core orifices must be l.

adjusted from an equal temperature configuration to an equal l

flow configuration, which requires approximately 4 to 6 l

hours.

Orifice adjustments are continuously performed

'l during a power reduction and the change in configuration is l

initiated about 12-14% power.

In addition, the auxiliary l

boiler (s) is brought on-line to provide sufficient drive l

capability for the helium circulators when adequate nuclear l

generated steam 's not available (approximately 8% power).

l The shutdown requirements of LCO 3.0.3 do not apply in l

SHUTDOWN and REFUELING because the ACTION requirements of l

individual specifications define the remedial measures to be l

taken.

s b

n Fort-St. Vrain il 1

'u 4

Technical Specifications Amendment No. 77 J'

Page 3/4 0 +

4.

.l BASIS FOR SPECIFICATION LCO 3.0/SR 4.0 (Continued) l 3,0.4 This specification provides that entry into an OPERATIONAL l

MODE or other specified applicability-conditions must be l

made with:

(1) the full compliment'of required systems,.

l equi pment', or components OPERABLE and (2) all other.

l parameters as specified.in the Limiting. Condition for l

Operation being met without regard for allowable deviations

{

l and out-of-service provisions ~ contained in the ACTION I

statements.

l The purpose of this specification is to ensure that faci'.ity I

operation is not initiated or that higher MODES-of operation l-are not entered when corrective action is being taken to l

obtain compliance with a specification by restoring l

l equipment to OPERABLE -status or. parameters to specified I

limits.. Compliance with ACTION requirements that permit l

continued operation of the facility for an unlimited period l

of tima provides an acceptable level of safety for' continued I

operation without regard to the status of the plant before l

'or after a MODE change. Therefore, in this case, entry into l

an OPERATIONAL MODE or other specified condition may be made j

in accordance with the provisions

.of the ACTION I

requirements.

The provisions of this' specification-should I

not, however, be interpreted as. endorsing the ' failure to l

exercise good practice in restoring systems or components to L

l' OPERABLE status before plant startup.

l l

When-a shutdown is required to comply with ACTION l

requirements, the provisions of-LCO 3.0.4 do not apply l

because they would delay placing the facility in a lower l

l MODE of operation.

l Exceptions to this provision have been provided for a p

l limited number of specifications when startup with l-l inoperable equipment would not affect plant safety. These L

l-exceptions are stated in the ACTION statements of the f

l appropriate specifications.

l 3.0.5 The CALCULATED BULK CORE TEMPERATURE (CBCT) is used in the l-FSV Technical Specification as an indicator of decay heat c

l levels that determines the applicability of LCO or ACTION E

l requirements.

LC0 4.0.4 describes the procedure for D

l calculating Bulk Core Temperature.

Refer to this l

Specification and its Basis.

Fort St. Vra'n 41 e.K %.

Technical. Specifications Amendment No,177

L,

Page 3/4 0 i

~

_l-BASIS FOR SPECIFICATION LCO 3.0/SR 4.0 (Continued)

,l 4.0.1 The ~ Surveillance Requirements specified in these Technical y

l

. Specifications define the

tests, calibrations,-

and

}

l

-inspections which ensure the performance and OPERABILITY of

-l' equipment essential to safety o r.

equipment _ required; to

'l prevent or mitigate the consequences of abnormal situations..

l-These

. requirements are based on. the-Surveillance l~

Requirements stated in the Code of Federal Regulations, 10 l

CFR 50.36(c)(3):

q l

" Surveillance Requirements are requirements relating to l

test, calibration, or inspection to ensure that- 'the j

lf necessary quality of systems and components is maintained, i

l that facility operation will be within safety limits, and l'

that the limiting conditions of' operation will be met."

l This specification. provides that surveillance activities l-necessary to ensure that the Limiting-Conditions for a

l Operation are being met and that they will be performed i

during the OPERATIONAL MODES or other conditions for' which r

l the Limiting Conditions for Operation are applicable.

-l-Surveillance Requirements do not have-to be performed when

~

l-the facility.is in an OPERATIONAL MODE for which the l

requirements of the associated Limiting Condition for l

Operation do not apply unless otherwise specified.

b l

Provisions for additional surveillance activities to be.

1 performed without regard to the applicable OPERATIONAL MODES

-l.

or other conditions are provided 'in the individual i

l Surveillance Requirements.

Surveillance Requirements for l

l Special Test Exceptions need only be performed when the L

l Special Test Exception is being utilized as an exception to

'?

l

'an individual specification, t

i l 4.0.2 The provisions of this specification provide ~ allowable l

to.lerances for performing surveillance activities beyond l

those specified in the nominal SURVEILLANCE INTERVAL. These l

tolerances are necessary to provide operational flexibility l

because of scheduling and performance considerations.

The l

phrase "at least" associated with a surveillance frequency l

does not negate this allowable tolerance value and permits j

the performance of more frequent surveillance activities.

i.

h Fort St, Vrain #'.

rR",

Technical Specifications-

-Amendment No. 77 Page 3/4 0_9' l BASIS FOR SPECIFICATION LCO 3.0/SR 4.0 (Continued) l Specification 4.0.2 establishes the limit for.which the l

'specified time interval for Surveillance Requirements may be' l

• extended.

It permits an allowable extension of the normal l

surveillance interval to' facilitate' surveillance scheduling

' l and consideration of plant operating conditions that may not-l be-suitable for conducting the surveillance; e.g., transient l

conditions or other ongoing surveillance or maintenance l

activities.

It also provides flexibility to accommodate the I

length of a fuel cycle for surveillances that are performed

' l at each refueling outage and are specified with an 18 month l

surveillance interval.

.It is not intended that' this-l provision be used repeatedly as a-convenience to extend I

surveillance intervals beyond that specified for 1

-surveillances.that are not performed during refueling i

outages.

The limitation'of Specification 4.0.2 is based on i

engineering judgement and the recognition that the most i

probable result of any particular surveillance being i

performed is the verification of-conformance with the l

Surveillance Requirements. This provision is sufficient to I

ensure that the reliability ensured -through surveillance I

activities is not significantly degraded beyond that l.

obtained from the specified surveillance interval.

l 4.0.3 The. provisions of this specification set forth the criteria l

for determination of compliance with the OPERABILITY l

requirements of the Limiting Conditions for Operation.

l Under these criteria', equipment,: systems, or components are l

assumed,to be OPERABLE if the associated surveillance I

activities have been satisfactorily performed within the~

l specified time interval. Nothing in this provision is to be I

construed as defining equipment, systems or components l

OPERABLE when such items are found or known to be inoperable I

although still meeting the Surveillance Requirements.

___b______._.__.________

._._._._____1

b FertiSt. Vrain 11.

-l W

Technical' Specifications'-

y.

Amendment.No; 77 H

Page 3/4 0-10 1

p l BASIS FOR' SPECIFICATION LCO 3.0/SR 4.0 (Continued)__

I e'

.l This specification also clarifies that the --- ACTION

.l' requirements are applicable when Surveillance Requirements

-l have not been completed within the allrwed SURVEILLANCE e

l INTERVAL and that the time limits of the ACTION requirements c

l apply from the point in time it is identified that a I

surveillance has not been performed and not at the time that.

t I

the allowed SURVEILLANCE INTERVAL was exceeded.

Completion 1

of the Surveillance Requirement within the allowable outage

'l time limits of the ACTION requirements restores compliance-l with the requirements of SR 4.0.3.

However, this does not l

negate the fact that the failure to have performed the l

surveillance within the allowed ' SURVEILLANCE

INTERVAL,

.I defined by SR 4.0.2, was a violation of the OPERABILITY 1

-l requirements of a Limiting Condition for Operation that is l

subject.

to possible enforcement action.

Further,' the l

failure to perform a surveillance per SR 4.0.2 is a l

violation of a Technical Specification requirement and is, I

therefore, a reportable event under the requirements of 10 l

CFR 50.73(a)(2)(1)(B) because it is a condition prohibited l

by the plant's. Technical Specifications, r

l If the allowable outage time limits of the ACTION l

recuirements are less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or a shutdown is l-required to comply with ACTION requirements, 6.g., SR 3.0.3,

'I

t. 24-hour allowance is provided to permit a delay in l

implementing the ACTION requirements.

This provides an

_l adequate time limit to complete Surveillance Requirements l

that have not been performed.. The purpose of this allowance l

1s to permit the completion of a surveillance before a I

shutdown is required to comply with ACTION requirements or l

before other remedial measures would be required that may l

preclude completion of a surveillance. The basis for this l

allowance includes consideration for plant conditions, I

adequate planning, availability of personnel, the time l

l required to perform the surveillance, and the safety I

significance of the delay in completing the required l~

I surveillance.

This provision also provides a time limit for i

the completNn of Surveillance Requirements that become l

applicable as a consequence of MODE changes. imposed by

'~

l-ACTION requirements and for completing Surveillance L

l Requirements that are applicable when an exception to the I

requirements of SR 4.0.4 is allowed, unless a longer l

exception is specifically allowed.

If a surveillance is not l

I completed within the 24-hour allowance, the time limits of I

the ACTION requirements are applicable at that time. When a l

surveillance is performed within the 24-hour allowance and l

the Surveillance Requirements are not met, the time limits l

of the ACTION requirements are applicable at the time that l

the surveillance is terminated.

cet St. Vrain al Technical Specifications Arenoment No. 77 Page 3/4 0-11 l BASIS FOR SPECIFICATION LCO 3.0/SR 4.0 (Continued) i Surveillance Requirements do not have to be performed on l

inoperable equipment because the ACTION requirements de fine l

the remedial measures that apply. However, the Surveillance

(

Requirements have to be met to demonstrate that inoperable i

equipment has been restored to OPERABLE status.

l 4.0.4 This specification ensures that the surveillance activities l

associated with a Limiting Condition for Operation have been l

performed within the specified time interval prior to entry l

into an OPERA 1IONAL MODE or other applicable condition.

Tne l

intent of this provision is to ensure that surveillance l

activities have been satisfactorily demonstrated on a

l current basis as required to meet the OPERABILITY l

requirements of the Limiting Condition for Operation.

l Under the terms of this specification, for example, dur8.ng l

initial plant STARTUP or following extended plant outages, l

the applicable surveillance activities must be performed I

within the stated SURVEILLANCE INTERVAL prior to placing or l

returning the system or equipment into OPERABLE status.

p l

When a

shutdown is required to comply with ACTION

)

requirements, the provisions of SR 4.0.4 do not apply l

because this would delay placing the facility in a lower l

MODE of operation.

l l

9 e

l I

Fort St. Vrain 01 i

Technic 6 Specifications i

Amendment No. 77

)

Page 3/4 1-1 l

1 9

i l REACTIVITY CONTROL l 3/4.1.1 CONTROL ROD PAIR OPERABILITY l LIMITING CONDITION FOR OPERATION i

i t

l 3.1.1 All control rod pairs not fully inserted shall be OPERABLE l

with:

l A.

A scram time lest than or equal to 152 seconds from the l

fully withdrawn position, t

l B.

A control rod drive (CRD) motor temperature less than or i

1 equal to 250 degrees F, l

C.

A heliur.: purge flow not carrying condensed water to each l

CRD penetration when reactor pressure is above 100 psis, l

and l

3.

The absence of a slack cable alarm.

l APPLICABILITY:

POWER, LOW POWER, and STARTUP l ACTION:

l A.

With one or more control rod pairs inoperable due to l

being immovable (i.e.,

not capable of being fully l

inserted), within 10 minutes initiate a reactor shutdown l

and an assessment of the SHUTOOWN MARGIN, and be in at l

1 east 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 B.

With one control rod pair inoperab'le due to having a l

Scram time greater than 152 seconds, opera tion may j.

continue provided that within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s:

l 1.

The control rod pair is restored to OPERABLE status, l

or l

2.

The control rod pair is fully inserted, or l

3.

The SHUTDOWN MARGIN requirement of LCO 3.1.4 is l

satisfied with the control rod pair considered l

inoperable in its present position.

l If none of the above conditions can be met, be in at i

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

a

--e m,.

N

~

Foet it. V*a 9 01' Technical Specifications Amendment No. 77 Page 3/4 1-2 I

r l SPECIFICATION LCO 3.1.1 (Continued)

L l

C.

With two or more control rod pairs inoperable due to L

l having a scram time greater than 152 seconds, within 10 l

l minutes initiate a reactor shutdown and be in at least l

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

l D.

With one or more control rod pairs having a CRD motor I

temperature greater than 250 degrees F,

operation may l

continue provided that within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s:

l 1.

The control rod pair (s) is restored to OPERABLE l

status, or l

2.

Surveillance testing per SR 4.1.1.A is performed on l

l the control rod pair (s) once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the l

CRD moter temperature exceeds 250 degrees F.

With l

t;ne or more control rod pairs exceeding a scram time l

of 152 seconds, comply with ACTIONS B or C above, i

With scram times less than or equal to 152 seconds, l

up to four control rod pairs with CRD motor I

temperatures greater then 250 degrees F may be l

considered OPERABLE for SHUTDOWN MARGIN l

determination.

l E.

With no purge flow to one CRD penetration, operation may I

l continue provided that within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s:

l 1.

Purge flow is restored to the CRD penetration, or l

2.

The control rod pair is fully inserted, or l

3.

The SHUTDOWN MARGIN requirement of Specification l

3.1.4 is satisfied with the control rod pair l

considered inoperable in its present position.

l If one of the above conditions cannot be met, be in at l

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

l F.

With no purge flow to two or more CRD penetrations:

l' 1.

Restore purge flow within 2 hou'rs, or l

2.

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

b

h

~

w Fort St. Vrair 01 i~.

a.

Technical Specifications Amendment No. 77 4

Page 3/4 1-3

)

i l

l SPECIFICATION LCO 3.1.1 (Continued) l G.

With the water level in the knock-out pot for the CRD l

purge flow lines greater than 6 inches, but with the i

l knock-out pot not flooded:

l 1,

Within I hour dra.in the knock-out pot and establish

(

l a helium purge flow not carrying condensed water, or 1

l 2.

Be in at least 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 H.

With the knock-out pot for the CRD purge flow lines t

I flooded:

i 1.

Be in at least 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

and l

2.

Perform surveillance SR 4.1.9.E.

t l

1.

With a slack cable alarm, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> determine l

whether a slack cable condition exists (i.e.,

a parted I

cable, detached cable, or failed instrumentation that is.

1 inaccessible for repair during operation).

If an actual l

slack cable condition exists, be in at least SHUTDOWN l

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

If the alarm is due to some l

other condition, restore the alarm to OPERABLE status l

within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or declare the affected control l

rod pair inoperable and comply with the requirements of l

Actier. B.

t l

J.

The provisions of Specification 3.0.4 are not applicable l

for changes between STARTUP, LOW POWER, and POWER.

l Prior to entry into STARTUP from SHUTDOWN, all l

requirements of this LCO must be met, without reliance l

on provisions contained in the ACTION statements.

Fort St. Vrain c1 c

Technical Specifications Amendment No. 77 page 3/4 1-4 l SURVEILLANCE REQUIREMENTS =_

l 4.1.1 Ea " ontrol rod pair shall be demonsteated OPERABLE:

l A.

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

l 1.

Verifying that all CRD motor temperatures are less l

than or equal to 250 degrees F.

j l

4.

k'ith one or more CRD motor tempeiature(s) l exceeding 215 degrees F:

l

1) The temperature of any CRD motor exceeding l

215 degrees F shall be recorded, l

2) A partial scram test as described in l

SR 4.1.1.8 shall be performed at least once I

per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on the control rod pair with l

the highest motor temparature and for all I

control red pairs greater than 250 degrees l

F, and l

3) A report on the partial scram test results l

and the maximum daily temperature of any l

control rod pairs with motor temperatures l

exceeding 215 degrees F shall be submitted l

to the NRC once every 31 days.

b.

If CRD motor temperature instrumentation is l

l inoperable, an engineering evaluation-shall be l

performed to determine CR0 motor temperature by l

comparison.

l 2.

Verifying purge flow to each CRD by verifying flow l

in each subheader, when reactor pressure is above l

100 psia; and l

3.

Verifying that the purge flow is not carrying l

condensed water by verifying that the water level in l

the knock-out pot is less than 6 inches.

l 4.

Verifying that the slack cable alarm is not l

actuated.

t a

6

6crt St. Vrade o1 Technical Specifications A;enoment No. 77 Page 3/4 1-5 r

l SPECIFICATION SR 4.1.1 (Continued) l-B.

Once per 7 days by:

l 1.

Performing a partial scram test of at least 10 l

inches on all partially $nserted and fully withdrawn l

control rod pairs except the regulating rod pair, i

l and verifying that the extrapolated scram time is l

1ess than or equal to 152 seconds; and

-l 2.

Performing a partial scram test of approximately 2 t

l inches on the regulating rod pair and veri fying l

control rod pair movement.

l C.

Prior to withdrawal of control rod pair (s) if not l

performed in the previous 7 days:

l 1.

By performing a partial scram test of at least 10 l

inches and verifying that the control rod pair i

inserts freely and can be considered scrammable, or l

2.

If withdrawai is to achieve criticality by l

performing a partial scram test of at least 10 l

inches on all control rod pairs being withdrawn, and

[

l verifying that the extrapolated scram time is less I

than or equal to 152 seconds.

I l

D.

During each shutdown of.10 days or longer (if not l

performed during the previous 31 days) by performing a l

full stroke scram test on all control rod pairs and l

verifying a scram time less than or equal to 152 l

seconds.

l E.

Following any maintenance on a CRD mechanism which could l

affect the control rod pair scram time, by performing a e

l l

full stroke scram test and verifying a scram time of I

l 1ess than or equal to 152 seconds.

I F.

Once per 18 months:

l I

1.

By performing a CHANNEL CALIBRATION and a CHANNEL i

l FUNCTIONAL TEST of the eight subheader CRD purge I

flow measurement channels, l

2.

By performing a CHANNEL FUNCTIONAL TEST of the CRD l

motor temperature and cavity temperature l

instrumentation, l

E

. Feet St. Veadn *!

y

-Technical Specifications Amendment No. 77 Page 3/4 1-6 I

l'SFECIFICATIO4 SR 4.1.1 (Continued)__,,_

t l

3.

By performing preventive maintenance on each CRD in l

4 scheduled sequence such that none of the drives l

l installed in the reactor will have gone more than 6 l

REFUELING CYCLES without receiving preventive i

l maintenance.

Dur.ing these 6 REFUELING CYCLES, no l

CR0 shall be in regulating rod pair service (without l

receiving preventive maintenance) for more than one l

REFUELING CYCLE.

The preventive maintenance shall l

l consist of inspecting and replacing as necessary'the i

l CR0 gears,_ bearings, brake pads,

cables, and j

l position instrumentation, and 4

l 4

By performing a CHANNEL CALIBRATION of the CRD motor l

and cavity temperature instrumentation on those CRDs I

undergoing preventive maintenance as described in 1.

SR 4.1 1.t.3 above.

i e

[

t b

)

e

7 Fo-t St. Vrain el y

Technical Specifications Amendment No. 77 l.

Page 3/4 1-7 1

b i

i l BASIS FOR SPECIFICATION LCO 3.1.1/SR 4.1.1 i

l Control rod pair OPERABILITY ensures that a minimum SHUTDOWN MARGIN l is capable of being maintained.

l l The control rod pair withdrawal accident analyses described in FSAR l

l Sections 14.2.2.6 and 14.2.2.7 were performec assuming a scram l insertion time of 152 seconds and a ramp reactivity insertion of 10.080 rielta k and 0.058 delta K, respectively.

l l Requiring the scram time to be less than or equal to 152 seconds will i ensure that the ramp reactivity rate is consistent with that assumed l in the accident analyses.

The full insertion scram time can be I determined either directly from a full insertion scram test or l indirectly from a partial scram test of 10 inches or more.

For the I partial scram test, the estimate of an extrapolated scram time of I less than or equal to 152 seconds is always based on assuming a scram t

l from the fully withdrawn position and not from the actual rod l position.

l The total calculated reactivity worth of'all 37 control rod pairs is l 0.210 delta k,

which is, significantly greater than the scram l reactivities assumed in the accident analyses.

Therefore, a single j control rod pair with a scram time greater than 152 seconds, as l allowed in ACTION B of the specifications, will have no impact o't the l calculated consequences of the control rod pair withdrawal accident.

i l Temperature Limitation l Control Rod Drive Mechanism (CRDM) qualification tests were performed l in a 180 degree F helium environment.

The motor and brake were l

l energized and deenergized in severe duty cycles up to once every 5 l seconds for 630,000 jog cycles and 5000 scrams of the CROM.

CRDM I motor temperatures ranged from 200 degrees F to 230 degrees F with an l=

l average of 215 degrees F during these tests.

During power ascension l testing, CRDM temperatures up to 213 degrees F were experienced at l power levels up to 70%. Using data obtained during power ascension l testing, a CRDM temperature of 260 degrees F was predicted for 100*;

l power conditions with an orifice valve fully closed.

The minimum l predicted open position for an orifice valve at 100*4 power is about l 10%, for which the predicted CRDM temperature is 250 degrees F.

l Tes'ts conducted to 100% power indicated these predictions to be l

l conservative because the maximum measured CRDM mot'or temperature was i

l 218 degrees F.

The operating temperature of the CRDM is limited by l the motor insulation which is derated to 272 degrees F to account for j motor temperature rise, frictional torque increase, and winding life l expectancy. See Section 3.8.1.1 of the FSAR.

r:

Fert St. Vrain M

+..

Technical Specifications Amencmeat No. 77 Page 3/4 1-8 L

l BASIS FOR SPECIFICATION LCO 3.1.1/$R 4.1.1 l CRDM motor temperatures are monitored to verify that they are less l than or equal to 250 degrees F.

CRDM motor temperatures are alarmed I at 215 and 250 degrees F, and are recorded on a multi-point recorder l when they, exceed 215 degrees F (FSAR 3.8.1.1).

This recorder F

l provides frequent monitoring (at.least one reading per minute) and l the data is retrievable as required.

Any. CRDM with a motor i

1 temperature greater than 215 degrees F shall be reenrded every 24

.l hours to document that the temperature is less than 250 degrees F.

l In addition, the partial scram test frequency is increased from once l per 7 days to once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on the control rod pair with the I highest motor temperature.

A partial scram test will be performed l once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on all control rod pairs with motor temperatures I exceeding 250 degrees F, to verify that the extrapolated scram time I is less than or equ61 to 152 seconds. Verifying a control rod pair i extrapolated scram time of less than or eaual to 152 seconds, will i ensure CRDM reliability with a motor temperature greater than 250 l degrees F.

These surveillances ensure that CRDM motor temperatures l exceeding 250 degrees F de not degrade the CRDM's reliability to l perform its design function when required. and up to four of these l control rod pairs may be considered OPERABLE in SHUTDOWN MARGIN J determination.

1 If the CRDM motor temperature instrumentation is inoperable, an I engineering evaluation will be completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from the I time tSe instrumentation is found to be inoperabic, to verify that l the CRDM motor temperature is currently less than 250 degrees F.

l Additional temperature instrumentation located on the underside l primary closure plate and the orifice valve motor plate can be used l to infer-the CRDM motor temperature by comparing these temperatures l with those on another CRDM in a similar region. Other factors such l as orifice valve position and historical temperature data may be used l to determine CRDM motor temperatures by comparison.

l Purge Flow l The purge flow into the CR0 assembly limits the upward flow of I contrainated primary system helium coolant.

Purge flow to each CRD l

l penetration is ensured by verifying that purge flow is maintained to l each subheader and by sealing all the valves between the subheaders l and the CRD penetrations in an open position.

j

~

Fort St:. Vra19 01

.c Technical Specifications Amendment No. 77 Page 3/4 1-9

{

l 4

l BASIS FOR SPECIFICATION LCO 3.1.1/SR 4.1.1 (Continued)_

l lA knock-cut pot, moisture element, and pressure transmitter are I

l ir. stalled in the CRD purge line, between the purified helium header l and the CRD purge flow valve (FSAR 3.8.1.1).

Just before the knock-i i

l-l out pot, an independent source of dry helium is connectible in the l event the purified helium header becomes unavailable. The pressure

( in the helium header will be maintained above reactor pressure.

The l knock-out pot reduces the probability of moisture in the helium purge l header entering the CRD penetrations by trapping any entrained water l in the helium.

An alarm indicates that water is collecting in the l pot.

l The loss of purge flow to any CR0 assembly could result in elevated l CRDM temperatures that would require the appropriate monitoring and l its associated partial scram testing.

l The knock-out pot is approximately 10 inches deep.

Verifying that, l the water level in the knock-out pots is less than 6 inches once I every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ensures that the helium purge flow is not carrying j

condensed water.

l Slack Cable Alarm l The tension of the cables supporting each pair of control rods is l monitorad by means of slack cable sensing switches (FSAR section l 7.2.2.2).

A slack cable alarm for a region may indicate a control l red-stuck in the guide channels of the core, a parted control rod l cable, ' dropped control rod aosorber sections, or a failure of the l alarm instrumentation.

There are provisions to allow limited motion l of the affected control rods (up to 3 inches) to determine whether a l rod or cable is stuck or a cable is broken, and various diagnostic l

[ techniques can be used to determine the OPERABILITY of the l-l instrumentation.

l Actions l

l The ACTION to initiate a reactor shutdown within 10 minutes if one or l

l more' control red pairs are inoperable due to being immovable (e.g.,

l l resulting from excess friction or mechanical interference) is l implemented because the cause of the problem may be indicative of a l generic control rod pair prbblem which may affect the ability to l safely shut down the reactor.' When ACTIONS are to be taken within 10 l minutes, no restoration is intended.

The ACTION should be taken I without delay and in accordance with established procedures.

l I

6 cert St. Vrain el e

- o, Technical Specifications Amendment No. 77 4-Page 3/4 1-10 l BASIS FOR $PECIFICATION LCO 3.1.1/SR 4.1.1 (Continued) l The ACTIONS providing for continued operation with one control rod l pair inoperable-due to causes other than being immovable are less l restrictive because the $HUTDOWN MARGIN can be met with the highest l worth control rod pair fully withdrawn (FSAR Section 3.5.3).

,l Continued operation with CRD motor temperatures greater than 250

~l degrees F is acceptable provided continued OPERABILITY is l demonstrated once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by partial scram tests.

l Because the SHUTDOWN MARGIN can be met with the highest worth control l rod pair fully withdrawn, an exception to LCO 3.0.4 (which prevents l moving up to a higher OPERATIONAL MODE while in an ACTION statement)

I can be made in this case.

l If purge flow is not maintained to two or more CRD penetrations, 2 l hours is provided to restore purge flow to the penetrations.

Tnis I restoration time will provide time to change out a helium bottle or I clear any blockage in the subheader, in order to restore purge flow l to the CRD penetrations.

Degradation of the CRD assembly due to lack

~

l of purge flow is a long term ef fect, and will not occur over a short i period of time.

s' lack cable alarm is received, an actual slack cable condition l If a l must be confirmed or ruled out within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. An immovable control l red pair is subject to the SHUTDOWN requirements of Action A and is l not considered a slack catale condition in Action I,, For the l identified slack cab'le conditions, the affected control rod absorber l sections would be inserted into the core or else unaffected as in the l case of an instrumentati'on problem.

The consequences of these l conditions are conservative, the condition is local, and a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> l determination time is acceptable.

In the event of an actual slack l cable condition, the ACTION to shut down in a controlled manner is l acceptable since this is indicative of a local and not generic l problem and since determination and resolution will require the l removal of that CRDM assembly from the reactor.

i l Surveillances l The regulating rod pair is the only control rod pair with automatic l response capability to a change in flux and is used to offset the l negative effects of partial scram tests performed on other control l rod pairs. A partial scram test of 2 inches on the regulating rod l pair does not induce unacceptable power transients and demonstrates j that the control rod pair is moveable.

Fort St. Vrafe *!

I Technical $pecifications i

Amendment No. 77 Page 3/4 1-11 I

i s

l BASIS FOR SPECIFICATION LCO 3.1.1/SR 4.1.1 (Continued)_.

i j Performing a partial scram test prior to achieving criticality l ensures control rod pair OPERABILITY prior to entering into a higher l

l OPERATIONAL MODE.

The full stroke scram test performed during each I

i l shutdown is the most accurate method of determining the scram time i because the actual scram time is measured over the whole length of l the control rod pair versus being extrapolated from a partial l

l distance.

l For control rod pairs that are withdrawn later in the operational i

l schedule, a partial scram test prior to withdrawal is performed to j

l ensure scrammability; no extrapolated scram time is determined. A 10 J

l inch withdrawal from the fully inserted position does not produce I consistently meaningful extrapolated scram times due to the CRD j mechanism inertia, but it does establish ease of movement.

l Performing this 10 inch test also minimizes the compensating l movements of the regulating rod that would be required for partial 1

l scram tests from greater distances.

l' Performing a full-stroke scram test following any CRD maintenance i

l ensures that the OPERABILITY and scram time of the control rod pair l was not affected by the maintenance.

l The specified CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST assures W

l that the instrumentation monitoring the eight subheaders providing I purge flow to the control rod drive penetrations is OPERABLE and 1.oss.

7 l of purge flew is detectable.

l The specified CHANNEL FUNCTIONAL TEST of the CRD motor temperature l and cavity temperature instrumentation will assure that the l instrumentation monitoring the CRD temperatures is OPERABLE and l capable of detecting any increase in CRD motor temperature.

l l The preventive maintenance program performed on those CRDs replaced l each REFUELING CYCLE ensures that by inspecting and replacing as l

l necessary any degraded parts the potential for CRD failure is l significantly reduced. Since the regulating rod pair CRD is used l

l more than any other CRD, it will be substituted with ancther CRD l after each REFUELING.

e

% rt St. Vrain c1

.,I0 Technical Specifications Amendment.No.

77 Page 3/4 1-12 i

1 REACTIVITY CONTROL i

l 3/4.1.2 CONTROL ROD PAIR POSITION INDICATION SYSTEMS - OPERATING j

r L

l i

l l LIMITING CONDITION FOR OPERATION L

1 3.1.2 The position indication instrumentation listed in Table l

3.1.2-1 for each control rod pair shall be OPERABLE and l

capable of determining control rod pair position within 10 l

inches.

l APPLICABILITY:

POWER, LOW POWER, and STARTUP l ACTION:

As shown in Table 3.1.2-1 i

l SURVEILLANCE REQUIREMENTS l 4.1.2 A.

Control rod pair position instrumentation OPERABILITY-l shall be verified by performing a CHANNEL CHECK on the I

control rod pair position instrumentation, as follows:

l 1.

Prior to withdrawal from the fully inserted l,

position, l

2.

Upon full withdrawal, and l

3.

Once per 7 days on all fully withdrawn, part' ally 1

inserted, and fully inserted control rod pairs l

except for fully inserted control rod pairs l

incapable of being withdrawn.

l a.

During. partial scram surveillances on fully l

withdrawn and partially inserted control rod l

pairs, the analog rod position indication shall l

be demonstrate.d OPERABLE by verifying that the l

change in analog indication is cont.istent with l

the direction of control rod pair travel.

The i

analog and digital position indications must I

agree within 10 inches of each other.

If a l

larger difference is observed, it shall be l

assumed that the analog indication is the l

inoperable channel, unless the analog indication I

can be proven to be accurate and OPERABLE by l

another means, and l

S I

P Fort St. Vrain 02 Technical Specifications

(

Amendment No. 77 L

Page 3/4 1-13 4

e L

l SPECIFICATI0N SR a.1.2 (Continued) t 1

l b.

During partial -scram surveillances on fully I

withdrawn control rod pairs, the rod-out limit l

indications shall be demonstrated OPERABLE by l

verifying that the rod-out indication clears l

when the control rod pair is inarted less than I

or equal to 6 inches and is on when the control l

red pair is fully withdrawn following the l

l-partial scram test.

i l

B.

Prior to each reactor startup and the first time during

-l or af ter startup when the control rod pair is withdrawn l

from the fully inserted position, the OPERABILITY of the t

l rod-in limit indication shall be verified for each I

control rod pair by:

l 1.

Verifying that the rod-in limit light is on, when l

the control rod pair is fully inserted, and

+

l 2.

Verifying that the rod-in limit light clears, when l

the control rod pair is withdrawn less than or equal l

to 6 inches.

l C.

Prior to each reactor start-up, and during the first l-outward motion of a control rod pair, the OPERABILITY of l

the analog and digital position indications shall be l

verified for cc:h control rod pair by:

l 1.

Verifying that the rod-in limit' light is on, when j

the control rod pair is fully inserted, and l

2.

Verifying that when the control rod pair is l

withdrawn a short distance, the rod-in limit light l

clears, when the analog and digital instrumentation l

indicates less than 6 inches.

1 If the analog and digital position indications l

indicate 6 or more inches, an engineering evaluation t

l shall be performed to determine the maximum l

l insertion limit for that control rod pair.

1

~

j f

L E

l

I Fort St. Vrain dl i-Technical Specifications i

Amendment No. 77

[E Page 3/4 1-14 L

I-l l

Table 3.1.2-1 l

CONTROL ROD PAIR POSITION INDICATION SYSfEMS l

l ll l

l MINIMUM POSITION l l

l l POSITION OF l POSITION INDICATION l INDICATION l

l l l CONTROL ROD l INSTRUMENTATION l INSTRUMENTATION l

l l l PAIR l SYSTEMS AVAILABLE l SYSTEMS OPERABLE I ACTION l

II I

I l

I l l Fully l Roc-in Limit.

l 1

l 1, 2 l

I l Inserted l Independent Means l

l l

j l'

l l l (Watt Meter Test) l l

l Il l

l l

l l l Partially l Rod-In Limit *,

l 2

l 1, 2 l

I l Inserted l Analog, Digital l

l l

Il l

I I

I l l Fully l Roc-out Limit, l

2 l

1, 2 l

l l Withdrawn l Analog, Digital l

l l

l l l

l l

l l

ACTION STATEMENTS l ACTION I With the nun.ber of OPERABLE position instrumentation systems l

less than the Minimum Position Indication Instrumentation l

Systems OPERABLE requirements, restore the required number l

of inoperable position indication system (s) to OPERABLE l

status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, or be in SHUT 00WN within the next 12 l

hours.

l ACTION 2 The provisions of LCO 3.0.4 are not applicable for changes l

between STARTUP, LOW POWER, and POWER.

Prior to entry into l-STARTUP from SHUTDOWN, all the requirements of the LCO must I

be met, without reliance on the provisions contained in the l

ACTION statement.

l l

• Demonstrated OPERABLE when last tested.

t

b rt St. Vrs** C b

Technical Specifications Amencment No. 77 Page 3/4 1-15 i BASIS FOR SPECIFICATION LCO 3.1.2/SR 4.1.2 e

l FSAR Section 7.2.2 assumes a long term misalignment of plus or minus 1

l 12 inches on control rod pair position to ensure an acceptable power l distribution for core burnep. This allows for a 2 foot separation l distance for the control rod pairs of any partially inserted shim l group.

Assuring a position accuracy of plus or minus 10 inches is i

l consistent with this misalignment allowance and provides for a 4 inch

! margin for operation when manually 9djusting the control rod pairs of l the shim group.

Each shim control rod pair is normally moved in l approximately 2. inch increments during operation to adjust the I regulating red pair to its mid operating position.

A 10 inch I position accuracy for all control rod pairs is also consistent with a f

l reactivity uncertainty of about 0.003 delta k,

which allows for I detecting core irregularities, such as an inadvertant release of l reserve shutdown material within a single core region.

Control red l pair withdrawal procedures require an evaluatien i f-the actual I critical control rod pair position-differs from the predicted I position during initial criticality by this reactivity worth.

l Control red pair position indication system OPERABILITY is required l to determine control red pair positions and to ensure compliance with I control rod pair alignment and position reouirements of LCO's 3.1.5 l and 3.1.6.

l Rod-out and red-in position indication is provioed by cam-actuated l switches. The cams are mounted on the same shaft as the rod po.sition l potentiometer. The shaft is directly coupled to a cable drum through l a gear train and rotates as required for the full length of control l rod pair travel. When a control rod pair is withdrawn from the fully l inserted position, the limit switch cams release the rod-in switch l causing the rod-in light to extinguish.

Rod position is transmitted l to the console by a potentiometer coupled directly to the drum l gearing.

The rod-in and rod-out limit switches and the rod position s

l potentiometer transmitters are duplicated to protect against the loss l of position indication.

l ACTIONS l If analog and rod-in limit indications are OPERABLE but digital l and/or rod-out limit indications are inoperable, operation may l continue.

Since both the analog and digital indications are taken l from the same shaf t and potentiometer, control rod pair position is I still capable of being established with only the analog indication.

l Rod-in limit indication capability is more critical than tod-out I limit indication for the purpose of determining SHUTDOWN MARGIN.

9 4

8 9

---

t

Fort St. Vrain el 4

Technical Specifications Amendment No. 77' Page 3/4 1-16 t

l BASIS FOR $PECIFICATION LCO 3.1.2/SR 4.1.2 (Continued)

I If the analog indication is inoperable, operation may continue with 1 one of the following conditions satisfied:

I a.

If the control rod pair is fully inserted, the position can be l

established by the rod-in limit. indication or verified by an l

independent means such as the watt-meter test.

Since the control l

rod pair is fully inserted, any other position indication is not l

requireo because its position of being fully inserted can be i.

l verified and used in the SHUTDOWN MARGIN calculation.

l b.

For the case when the control rod pair is partially inserted and l

the digital and rod-in limit indications OPERABLE. control red l

pair position can still be established by digital indication and l

if the control red pair were to be fully inserted its position l

could be verified.

Rod-in indication OPERABILITY is demonstrated l

when last tested.

l c.

For the case with the control rod pair fully withdrawn and rod-l out and rod-in limit indications OPERABLE, th:t control rod pair's l

position can be established (i.e.

fully withdrawn) or if the i

control rod pair were to be fully inserted, its position could be l

verified for the SHUTDOWN MARGIN calculation.

l If red-in limit indication were inoperable, operation may continue, l because a fully inserted control rod pair's position can be l established by an independent means such as the watt-meter test.

At l a partially inserted or fully withdrawn position, the control red l pair's pcsition can be determined by both digital and arralog l indications.

l If control rod pair position cannot be determined within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, l reactor shutdown is required within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

This ACTION l 1s required to satisfy the control rod pair worth and position l requirements of LCO 3.1.5, which prevents an unacceptable power l distribution.

l Surveillances

{ Control rod pair rasitit' indication instrumentation OPERABILITY is l verified by perfoNag a MNNEL CHECK before the control rod pair is l withdrawn from tnc fully inserted position, when it is fully l

l withdrawn, and once per 7 days.

This surveillance ensures position l indicat %n OPERABILITY prior to a reactor startup and during l operation.

l i

Fort St. V-air #1 Technical. Specifications t

Amendment No. 77 Page 3/4 1-17 t

f l BASIS FOR SPECIFICATION LCO 3.1.2/SR 4.1.2 (Continued) l During the partial scram test (once per 7 days during operation) l analog indication is verified OPERABLE by confirming that the change l in analog indication is consistent with the direction of control rod l pair travel.

l If a difference of greater than 10 inches exists between the analog j and digital position indications, the analog indication is considered l inoperable, unless proven accurate by another means.

The analog l indication may be proven to be accurate and OPERABLE by fully l inserting the control rod pair and verifying that the analog '

l indication is more accurate than the digital indication at the fully l inserted position as determined by the rod-in limit indication or the l watt-meter test.

l The rod-in limit indication is verified to be OPERATING at the fully l inserted position when the control rod pair is withdrawn a short I distance.

This surveillance ensures that a fully inserted control l rod pair's position can be established during operation by verifying l OPERABILITY of each control rod pair prior to each startup and also l when the control rod pair is first withdrawn from the fully inserted l position, t

l To ensure position indication is capable of being established at the l partially inserted to fully withdrawn position (during' operation) l both the analog and digital positions are verified OPERABLE at the l fully inserted position when the control rod pair is withdrawn a l short distance.

This surveillance is performed prior to startup or l during the first outward control rod pair motion.

l l The position indication potentiometers and associated coupling can be l

[ damaged by an overtravel of minus 6 inches. This damage is prevented I by initially requiring the control rod pair position indication to l

l indicate less than 6 inches when the rod-in limit indication clears l and then by procedurally preventing control rod pair insertion past l zero, even if rod-in limit indication is not received.

The l requirement for position indication to be less than 6 inches when I rod-in limit indication is received imposes an enhanced accuracy l requirement at this position.

The result is that since procedurally l the centrol rod pair is not inserted past a zero indication, and if l the position indication is within 6 inches of the actual position, l then the control rod pair will not be inserted beyond the minus 6 l inch damage limit, even if the rod-in limit instrumentation f ails.

l Since control rod pair position instrumentation cannot be I recalibrated without removing the CRD from the PCRV, the engineering i evaluation provides the necessary procedural controls to establish l

l individual control rod pair insertion limits for control rod pa. irs l whose position indications exceed 6 inches.

l

E Fort St. Venie 01 Technical Specifications F

Amendment No. 77 I

Page 3/4 1-18 i

I REACTIVITY CONTROL l 3/4.1.3 CONTROL ROD PAIR POSITION INDICATION SYSTEMS - SHUTDOWN l LIMITING CONDITION FOR OPERATION l 3.1.3 The position indication ' instrumentation shall be OPERABLE l

for each control rod pair in fueled regions capable nf being l

withdrawn and caoable of determining control rod pair 1

position within 12 inches with:

A.

A rod-out limit indication or analog or digital pos'ition I

l indication, when the control rod pair is fully l

withdrawn, or l

B.

A rod-in limit indication and either an analog or l

digital position indication, when the control rod pair l

1s fully inserted.

l APPLICABILITY:

SHUTDOWN and REFUELING l ACTION: 'With any of the above required position indication l

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

l A.

Restore the inoperable position indication l

instrumentation to OPERABLE status, or l

B.

Verify full insertion of the control rod pair by other l

independent means (e.g., watt-meter testing), or l

C.

Consider the control rod pair fully withdrawn and meet l

the SHUTDOWN MARGIN requirements of LCO 3.1.4.

l SURVEILLANCE REQUIREMENTS l'4.1.3 A.- Control rod pair position instrumentation required by l

LCO 3.1.3 shall be demonstrated OPERABLE by performing a l

CHANNEL CHECK as follows:

l l

1.

Prior to withdrawal from the fully inserted l

position, i

2.

Upon full withdrawal, y

f Fort St. '.'rair al Technical Specifications Amendment No. 77 Page 3/4 1-19 1

l SPECIFICATION SR 4.1.3 (Continued) _

l 3.

Once per 7 days on all control rod pairs except for l

fully inserted control rod pairs which are incapable i

l of being withdrawn, and l

8 4.

After an OPERATIONAL MODE change to SHUTDOWN from l

STARTUP.

j l

The analog and digital position indications shall be i

l within 12 inches of each other.

If a larger L

l difference is observed, it shall be assumed that the l

analog indication is the inoperatie channel, unless l

.the analog indication can be proven to be accuratt l

and OPERABLE by another means.

l B.

Once 'per 18 months for control rod pairs in fueled l

regions, a CHANNEL FUNCTIONAL TEST of each control rod l

' pair's. redundant "in" and "out" limit switches and l

l analog and digital rod position indication systems, l

shall be performed.

l C.

A CHANNEL CALIBRATION of the control rod pair redundant l

"in' and "out" limit switches, and the analog and l

digital rod position indication systems, shall be l

performed on all CRDs removed for repair or maintenance.

l 0.

When i n' REFUELING, prior to each control rod pair l

withdrawal (unless the surveillance has been performed l

within the previous 7 days) the OPERABILITY of the l

analog and digital position indications shall be l

verified for that control red pair by:

l 1.

Verifying that the rod-in limit light is on, when l

the control rod pair is fully inserted, and l

2.

Verifying that when the control rnd pair is l

withdrawn a short distance, the rod-in limit l

indication clears when the analog and digital l

instrumentation indicates less than 6 inches.

l If the analog and digital instrumentation indicates l

6 o more inches, an engineering evaluation shall be l

l performed to determine the maximum insertion limit l

for that control rod pair.

1, l

l l

l 1

Fort St. Vrain al Technical Specifications Amendment No. 77 l

Page 3/4 1-20 l BASIS FOR SPECIFICATION LCO 3.1.3/SR 4.1.3 i

l This specification involves control rod pairs that are either fully i

l inserted or fully withdrawn in fueled regions; therefore, the I accuracy requirements are different from those in LCO 3.1.2 for l operational considerations. The relative reactivity worth for the l total control rod pair bank as a function of withdrawal pontion is l given in FSAR Section 3.5.3 (Figure 3.5-2).

Experimental results on l control rod ' pair worth versus withdiawal position have indicated a l reduced worth for the first portion of control rod pair withdrawal l and has been substantiated with new analyses.

From this revised l

l calculated data and a calculated bank worth.of 0.210 delta k, it can l be shown that a reactivity uncertainty of 0.003 delta k results in I the total bank position uncertainty of 17 inches at full insertion l and 13 inches at full withdrawal.

The reactivity uncertainty of l 0.003 delta k is acceptable for the SHUTDOWN MARGIN and is consistent l with that used to detect core irregularities, such as occasions of l inadvertant release of reserve shutdowr. material within a single core i region. Control rod pair withdrawal procedures require an evaluation l if the actual critical control rod pair position differs from the l predicted position during the approach to criticality by the l reactivity worth of 0.003 delta k.

Verifying position accuracy l within 12 inches is consistent with these control red pair position l uncertainties.

l If position indication instrumentation is inopertble, a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> l ACTION time is alloweo because the SHUTDOWN MARGIN requirements have l been met prior to position indication instrumentation inoperability.

l Control rod pair position indication instrumentation OPERABILITY is 1 verified by performing a CHANNEL CHECK before the control rod pair is l withdrawn from the fully inserted position, when it is fully l withdrawn, once per 7 days, and after an OPERATIONAL MODE change to l SHUT 00WN from STARTUP on those control rod pairs capable of being l withdrawn.

The Basis for Specification 3/4.1.6 describes the methods l of making 'a control rod pair incapable of being withdrawn.

This l surveillance ensures position indication OPERABILITY when the reactor l 1s shutdown and during any refueling operations.

l Once per 18 months for control rod pairs in fueled regions, a CHANNEL l FUNCTIONAL TEST will be performed on the control red pair redundant l "in" and "out" limit switches and the analcg and digital rod position l indication systems.

This surveillance ensures that the entire l position indication system is OPERABLE prior to a reactor startup.

., ~ -

-a

Fort St. Vrain #1 Technical Specifications-Amendment No. 77 Page 3/4 1-21 l BASIS FOR SPECIFICATION LCO 3.1.3/SR 4.1.3 (Continued)

I In conjunction with CRD removal from the PCRV, a CHANNEL CALIBRATION i will be performed on the control red pair redundant "in" and "out" l limit switches, and the analog and digital rod position indication l systems. A CHANNEL CALIBRATION on the CRD position ihdication l instrumentation cannot be performed while the control rod pairs are l installed in the PCRV; therefore, a calibration is performed only on l the control rod pairs removed for repair or maintenance.

l During REFUELING, the rod-in limit indication, and analog and digital l indications will be verified OPERABLE (for those control rod pairs l capable of being withdrawn) within 7 days prior to control rod pair l withdrawal.

I Control rod pairs physically cannot be inserted into defueled l regions.

For this reason, this specification applies only to control I rod pairs in fueled regions.

4 s

l

Fcrt St. V-ain di t

Technical Specifications Amendment No. 77 Page 3/4 1-22 l REACTIVITY _ CONTROL l 3/4.1.4 SHUTDOWN MARGIN l LIMITING CONDITION FOR OPERATION l 3.1.4 The reactor SHUTDOWN MARGIN sna11 be greater than or equal l

to 0,01 delta k.

l APPLICABILITY: At all. times l ACTION:

l l

A.

When in POWER, LOW POWER, and STARTUP, with the SHUTDOWN l

l MARGIN less than 0.01 delta k:

~

r l

1.

Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, insert sufficient control red pairs l

to achieve the specified SHUTDOWN MARGIN, or l

2.

Be in at least 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 B.

When in SHUTDOWN, with the SHUTOOWN MARGIN less than 1

0,01' delta k, within I hour, either:

l 1,

Insert sufficient control rod pairs in fueled l

regions to achieve the specified SHUTDOWN KARGIN, or l

2.

Actuate sufficient reserve shutdown material in l

l l

fueled regions to achieve the specified SHUTDOWN l

MARGIN.

l C.

When in REFUELING, with the SHUTDOWN MARGIN less than l

0.01 delta k:

i l

1.

Immediately suspend all control rod pair or fuel l

manipulations involving positive reactivity changes.

l and l

2.

Within I hour either:

l a.

Fully insert sufficient control rod pairs into l

fueled regions to achieve the specified SHUTDOWN l

MARGIN, or l

b.

Actuate sufficient reserve shutdown material l

into fueled regions to achieve the specified l

SHUTDOWN MARGIN, 9

C

[

Fert St. Vrair. di l

- 4,-

Technical Specifications L

Amendment No. 77 Page 3/4 1-23 i

i

! SURVEILLANCE REQUIREMENTS i

I l 4.1. 4 -

SHUTDOWN MARGIN shall be assessed as follows:

l A.

When in POWER, LOW POWER, or STARTUP:

h l

l 1.

Once per_7 days, t

i 2.

In assessing -the SHUTDOWN MARGIN _ the following l

conditions shall be assumed:

l j

a.

The highest worth control rod pair is fully l

withdrawn, l

b.

All OPERABLE control red pairs are fully l

inserted with all inoperable control rod pairs l

in their pre-scram po>ition, I

c.

The CORE AVERAGE TEMPERATURE is equal to 220 l

degrees F, and l

d.

Full decay of Xe-135, no buildup of Sm-149, and l

no decay of Pa-233 beyond that present at I

shutdown.

l B.

When in SHUTDOWN:

l 1.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after each reactor shutdown when all I

control rod pairs cannot be verified fully inserted l

in fueled regions, 1

2.

Prior to control red pair withdrawal, if all control I

rod pairs are not fully inserted in fueled regions l

prior to withdrawal action, and l

3.

Prior to control rod pair withdrawal to achieve l'

criticality, to confirm that upon reaching l

criticality the SHUTDOWN MARGIN requirement can be l

met.

l 4.

In assessing the SHUTDOWN MARGIN the following l

conditions shall be assumed:

l a.

The highest worth control rod pair is fully l

withdrawn, l

b.

All OPERABLE control red pairs are fully 1

inserted in fueled regions and inoperable l

control rod pairs in their known position or l

fully withdrawn, o

~

~

Fort St. Vrain #1 l

-o, Technical Specifications t

Amendment No. 77 J+

Page 3/4 1-24 i

i l SPECIFICATION SR 4.1.4 (Continued) f l

. The CORE AVERAGE TEMPERATURE is equal to 80 1

degrees F, and l

d.' Full decay of Xe-135, full buildup of Sm-149, l

l and Pa-233 decay as a function of time after l

Shutdown.

l C.

When in REFUELING:

I 1.

Prior to control rod pair withdrawal, if all control l

rod pairs are not fully inserted into fueled regions l-prior to withdrawal action, and l

2.

Prior to the removal of the control rod pair in a

[

l region to be refueled or repaired.

l 3.

In assessing the SHUTDOWN MARGIN the following I

conditions shall te assumed:

l a.

The highest worth control rod pair capable of l

being withdrawn is ful'ly withdrawn, l

b.

Control rod pairs being withdrawn for

-l refueling / repair, SHUTDOWN MARGIN assessment, or l

OPERABILITY test purposes, are fully withdrawn, l

c.

All other OPERABLE control rod pairs are fully l

inserted into fueled regions and incapable of l

being withdrawn, I

d.

Inoperable control rod pairs are in their known I

position or fully withdrawn, l

e.

For planned CORE ALTERATIONS, the core shall be

)

in its most reactive configuration,

v-Fort St. Vraie. 01 Technical. Specifications Amendment No. 77

~

Page 3/4 1-25

[!

- lSPECIFICATIb~NSR4.1.4(Centinued)

I'

' l f.

The CORE AVERAGE TEMPERATURE is equal to 80 I

degrees F, and l

g:

Full decay of Xe-135, full buildup of Sm-149, I

and Pa-233 decay as a function. of time after l

shutdown.

l e

a e

\\

l l

l I

lc.

Fort St. Vra" ol i'

Technical Specifications A endment No. 77 Page 3/4 1-26

[

[

[

l BASIS FOR SPECIFICATION LCO 3.1.4/SR 4.1.4

~

i l A.

SHUTDOWN MARGIN - OPERATING l

The purpose of this LCO is to ensure that during operation a i

sufficient amount of negative reactivity in control rod pairs is l

capable of being inserted by the automatic and manual scram l

functions to shutdown the reactor with the highest worth control l

rod pair fully withdrawn.

A SHUTOOWN MARGIN of at least 0.01-I delta k has been specified at a CORE AVERAGE TEMPERATURE of 220 1

degrees F.

The CORE AVERAGE TEMPERATURE will normally be I

significantly above 220 degrees F for several days following a l

scram from power yielding a SHUTDOWN MARGIN greater than 0.01 l

delta k.

l Changes in the isotopic inventory following a reactor shutdown, l

of fission product pnisons Xe-135 and Sm-149, and heavy metal Pa-l

233, are also considered.

These changes are due to the buildup l

and decay of precursors as well as decay of their current

-l concentration.

For Xe-135, both the precursor and Xenon isotope I

decay in hours, with half-lives of 6 and 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> respectively, l

and consequently.Xe-135 initially builds up to a peak value in l

about 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,' and T. hen fully decays in a few days.

Since full

~

l decay occurs in a few days, it is conservatively assumed to be l

fully decayed at the time of shutdown. The precursor for Sm-149 l

has a half-life of a few days, while the decay of Sm-149 occurs l

over several years, so the buildup occurs over many days and is l

conservatively assumed to remain at its current value at l

shutdown.

The decay of Pa-233 to fissile U-233 occurs over a l

period of about 100 days, and it also is assumed to remain at its l

current value at shutdown.

.l Any control rod pair that is demonstrated CPERABLE per SR 4.1.1 L

l will be assumed to be fully inserted and any inoperable control-l l

rod pair will be assumed to be at its pre-scram position.

This l

is consistent with FSAR Section 3.5.3, which demonstrates that l

there is at least 0.014 delta k SHUTDOWN MARGIN with one control l

l rod pair fully withdrawn under any core condition in the l

l equilibrium core and larger for any core condition prior to the l

equilibrium core.

l l

l Assessment of the SHUTOOWN MARGIN requirements once per 7 days l

ensures that changes in the core reactivity as a result of burrup l

have not occu'rred which would make the previous verification I

invalid.

The core reactivity changes as a result of burnup occur l

slowly and a 7 day surveillance during operation is sufficient.

l In addition, the ACTION statements of LCO 3.1.1 require more l

frequent assessment.

if a control rod pair is determined l

l inoperable, or its exact position is uncertain.

r

\\

Fort St.' Vrain #1 e

Technical Specifications Amendment No. 77 i

Page 3/4 1-27

[

l BASIS FOR SPECIFICATION LCO 3.1.4/SR 4.1.4 (Continued) i l B.

SHUTDDWN and REFUELING l

The purpose of this specification is to ensure that during l

SHUTDOWN and REFUELING a sufficient number of control rod pairs l

I are fully inserted into fueled regions to keep the reactor in a l

shutdown condition. A SHUTDOWN MARGIN of at least 0.01 delta k I

l-has been specified at a CORE AVERAGE TEMPERATURE of 80 degrees F l

with decay of Xe-135, buildup of Sm-149 and some decay of Pa-233.

l The CORE AVERAGE TEMPERATURE will normally be significantly above I

80 degrees F for many months after shutdown, and the decay of Pa-l 233 occurs over

a. few months.

Therefore, the SHUTDOWN MARGIN l

immediately after achieving shutdown will normally be larger than I

the 0.01 delta k specified, and the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> delay in verification t

l of the SHUTDOWN MARGIN is sufficient for the p,urpose of this l

specification.

l This specification need only require that the control rod pair be l

actually inserted to r.chieve the specified SHUTDOWN MARGIN.

l Since full insertion can be verified by either rod position l

indication or another independent means, such as watt-meter

[

l testing per LCO 3.1.3, some additional time has been allowed.

l The Reserve Shutdown System was provided to ensure shutdown even l

in the event of failure to insert control rod pairs.

It is I

adequate to ensure shutdown even if all control rod pairs fail to l

insert (FSAR Section 3.5.3).

However, the contribution te *he l

SHUTDOWN MARGIN by the addition of reserve shutdown materia

.o l

a core region airsady containing an inserted control rod pair is i

minimal.

Therefore, it is sufficient to activate the reserve l

l l

shutdown material only in those regions whose control rod pairs

~

l are not fully inserted.

l For SHUTDOWN, the specified SHUTDOWN MARGIN assumes the full l

withdrawal of the highest worth control rod pair.

For REFUELING, l

l (which can include either fuel or control rod pair manipulations) l l

since all control rod pairs are disabled, except those involved l

with REFUELING per LCO 3.1.6, the requirement includes the l

addition of the highest worth control rod peir capable of being l

withdrawn in the SHUTDOWN MARGIN calculation.

Disabling of l

control rod drives by disabling the electrical supply to the l

drive motors or placing the reactor mode switch in the."off" l

position results in the inability to withdraw the control rod l

pair by action of the drive motor.

Therefore, the accidental l

withdrawal of any control rod pair in this manner does not have l

to be assumed in the SHUTDOWN MARGIN calculation.

1 g,

,a+

-.r..

m,.,

e

-*y-4*

-9'.;*-

=*-1.t 9

a 4

O

• /

Fcrt St. Vrain 81.-

(;* @i'?% fi-Technical Specifications 0,1 ;? -

Ar.endment No. 77 fs;F

• Page 3/4 1-28

}l$ 1 y _

b i-l-BASIS FOR~ SPECIFICATION LCO 3.1.4/SR 4.1.4 (Continued)

~

l The AdTION statement of. LCO 3.1.6, Control Rod Pair Position

-l ~

. Requirements-Shutdown, requires completion of the assessment of l

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

l Within the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown, the SHUTDOWN MARGIN is i

significantly larger than specified due to higher core l

temperatures and the presence of Xe-135 and Pa-233. A 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> l

delay will not compromise the validity of this specification.

l Assessment of the SHUTDOWN MARGIN prior to any control rod pair I

withdrawal, if all control rod pairs are not fully inserted, l

l prior to withdrawal to achieve criticality, and prior to removal l

of a control rod pair for refueling / repair, ensures that the l

requirements of this specification will-be met during these l

ACTIONS.

1 i

1 l

i i

i l

L

~

Fort.St. Vrain ci>

I l'"

Technical Specifications Amendment No. 77

~

Page 3/4 1-29.

F l REACTIVITY CONTROL

'l 3/4.1.5 CONTROL ROD PAIR POSITION AND WORTH REQUIREMENTS - OPERATING c

[. LIMITING CONDITION FOR OPERATION-I 3.1.5 Control rod pair position and worth requirements shall be as' l

follows:

l A.

Control rod pairs (except the regulating rod pair) shall

i l

be withdrawn or inserted in groups (three control red o

l pairs per group) except during scrams, control rod pair i

l l

runbacks, partial scram surveillance, or manipulations L

l-of additional.

control rod pairs-permitted by l

l Specification B.2 below.

l 4

l

. l-fully withdrawn except during partial scram testing and:

l B.

All. control rod pairs shall be either fully inserted or I

L l

1.

One shim group and the regulat,ing rod pair may be in

-l any position, and l

2.

Up to six additional control rod pairs may be l

inserted up to two feet.

l-C.

The maximum calculated control rod pair werth shall not l

exceed:

l 1.

0.047 delta k,

with the reactor critical at l

approximately 1.0 E-07 percent RATED THERMAL POWER l

(source power), and l

2.

At full power, that worth which would result in Rod l

Withdrawal Accident (RWA) consequences equal to l

those described for the worst case RWA in the AEC l

Safety Evaluation of Fart St.

Vrain dated January 1

20, 1972.

l APPLICABILITY:

POWER, LOW POWER, and STARTUP

r g

-m _

c Fort St. Veaia 61 Technical Specifications Amendment No. 77 P

Page 3/4 1-30 4

l'SPECIFICATIbNLCO3.~1.5(Continued) l ACTION:

l A.

With any control rod pair or group not in compliance l

with its position requirements either:

l 1.

Restore the control -rod pair (s) to an acceptable l

configuration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or 1

2.

Be in at least LOW POWER within tne next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, I

and SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, l

B.

With any control rod pair not in compliance with its l

worth limits, be in at least SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

l SURVEILLANCE REQUIREMENTS l 4.1.5-

~

l A.

Once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, each control rod pair position l

shall be verified to be in compliance with the above l

requirements.

l B.

At the beginning of each. REFUELING CYCLE, the reactivity l

I worth of-the control rod pair groups withdrawn from ' LOW l

POWER to POWER in the withdrawal sequence, shall be

.l measured.

The measured group worths shall be compared i

l with the calculated group worths to verify that the

.i I

calculated criteria upon which the selection of the-R l

control rod pair sequence was based has been satisfied.

j The measured group worth shall. agree with the calculated I

l group worth within plus or minus 20% for all groups I

l except groups 4A and 40, for which the measured group l

worth shall be _within plus 100%, minus 50% of the l

calculated group worth, i

9 9-a

g' 3, n Fc-t St. Vra b dl Technical Specifications 0 m Amendment No. 77 Page 3/4 1-31 P

3 l, BASIS FOR SPECIFICATION LCO 3.1.5/SR 4.1.5 l The :specif1 cation.of a control rod ~ pair withdrawal sequence and l position requirements during STARTUP and LOW POWER operation is

~..

l required to:

l a.

Assist in evaluating the reactivity, worth of control rod l

pairs withdrawn during the approach to criticality by l

ihuicated changes in the multipliec source neutrons, l

b.

Ensure that an acceptable power distribution is maintained I

(peaking factors within design limits) for the condition l

when many control rod pairs are still inserted, and 1

c.

Ensure that the calculated maximum worth control rod pair in l

STARTUP and LOW POWER if assumed. accidentally withdrawn, I

would result'in a transient'with consequences no more severe r

l than the control rod pair withdrawal accident (RWA) analyzed l

l in the FSAR.(Sections 3.5.3.1 and 14.2.2.7).

L L

l The specification of a, control rod pair withdrawal sequence and I position requirements during POWER are required to yield en l acceptable power distribution.

.In addition, the sequence ensures L

l that the combination of maximum single control rod pair worth and E

l available core temperature coefficients, in the event of an l-l accidental-control rod-pair withdrawal, will result in a transient l with consequences less severe than those analyzed in the FSAR.

The I RWA analyzed in the FSAR.is consistent with the RWA evaluation in the

[

l AEC Safety Evaluation of Fort St. Vrain dated January 20, 1972.

l l The maximum calculated control rod pair worth limit of 0.047 delta k l at approximately 1.0 E-7 percent power is based on the Maximum Worth l Control Rod Pair Withdrawal at Source Power analysis in FSAR l Section 14.2.2.7.

L 9

e:

-c.

Fort St. Vrain #1 Technical Specifications L

i,.

Amendment No. 77 Page 3/4 1-32 l BASIS FOR SPECIFICATION LCO 3.1.5/SR 4.1.5 (Continued) l The RWA analys's at rated power, as described in the FSAR, is based

.l on a maximum control rod pair worth of 0.012 delta k,

using l temperature coefficients equivalent to a reactivity defect from l refueling (220 degrees F) to operating temperature (1500 degrees F) l of 0.028 delta k.

For operation.in the range from 0 to 100 percent l power, the fuel temperature may be lower than the -full power l operating fuel temperature of 1500 degrees F.

This results in a l greater number of control rod pairs inserted for the critical l configuration, and a N rger maximum single control rod pair worth.

I In addition, since the temperature coefficients are greater at the l beginning -of the cycle, a single control rod pair worth as much as l 0.016 delta k is acceptable, i.e., the consequences of an RWA are i less severe (FSAR Section 14.2.1).

A value larger than 0.012 delta k I for a single control rod pair can be safely accommodated if fuel l temperatures are lower than ~1500 degrees F and/or the temperature i

l defect between refueling temperature (220 degrees F) and operating l temperature '(1500 degrees F) is greater than 0.028 delta k (FSAR j

l'Section 14.2.1.1).

l The presence of too many partially inserted control red pairs in the 1

l core will tend to push the flux into the bottom haif of the core and l

l raise the fuel temperatures.

The intra-region and axial power i peaking factors used in determining the control rod pair withdrawal l

l sequence for each REFUELING CYCLE will be maintained during normal L

l operation if the centrol rod pairs are inserted and withdrawn in j sequence and if partially inserted control rod pairs are limited as j

l noted above (FSAR Section 3.5.4).

L l The six additional control rod pairs which may be inserted up to 2 L

l feet into the core will permit the operator to move control rod pairs l'

. l to assist in regulating t,he core region outlet temperatures to those l specified in LCO's 4.1.7 and 4.1.9.

This has a minimal effect on the l axial. power distribution, resulting in an increase in the average I power density in the lower layer of fuel of less than 5%.

l The runback function inserts two pre-selected groups of three control o

l rod pairs during rapid load reduction (FSAR Section 7.2.1.2).

The K

l partial insertion of these control rod pairs, (FSAR Section 3.5.4.2) i l in addition to those noted above would increase the average axial I power peaking factor in the lower layer of fuel to about 0.85.

l b'egligible fuel particle kernel migration (SL 3.1) would occur with 1 this condition in the core for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

m

-I F

Fort St. Vrain el i

Technical Specifications j

Amendment No. 77'

.,b-Page 3/4 1-33 i

~

l BASIS FOR SPECIFICATION LCO 3.1.5/SR 4.1.5 (Continued) l The ACTION to be in at least LOW POWER within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and l SHUTDOWN in the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> requires an orderly shutdown to

l. reduce plant load and temperatures in a controlled manner.

Core l temperatures are significantly reduced as lower power levels are

~l reached, and in STARTUP negligible fuel particle kernel migretion l would occur as long as the ' minimum helium flow requirements (LCO l 4.1.9) are maintained.

l Verification of-control rod pair positions (by monitoring position l indication) once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ensures rod position changes with any l reactor power changes are noted, and is consistent with the l verification of INDIVIDUAL REFUELING REGION OUTLET TEMPERATURES once l per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (SR 5.1.8).

l The measurement of control rod pair group worths in the normal

_l withdrawal sequence at the beginning of each REFUELING CYCLE will I provide an ' evaluation of calculational methods in determining the l control rod' pair group worths in the core configuration for that I cycle.

The criteria used in selecting the control rod pair sequence l 1s based on calculated data for the maximum worth for any individual l control rod pair as well as the calculated peaking factors (region, l intra-region, and axial) in the normal operating ' control rod pair l configuration.

Since the core configuration changes for each l REFUELING CYCLE (a new segment includes approximately one sixth of l the total core) thi s evaluation confirms the ability to predict l control rod pair worths in that specific configuration.

l The acceptance criteria for the comparison of measured versus l calculated control rod pair group worth within plus or minus 20*4 l

l includes an allowance _for the calculated uncertainty of plus or minus E

l 101(FSAR Section 3.5.7.4) and uncertainty in the measurement.

A j larger acceptance criteria is needed for control rod pair groups 4A l

l and 4D because of a larger uncertainty in the calculated values.

)

l Groups 4A and' 40 are five column regions located at the core-I reflector interface, and the analytical model for control rod pair l worth calculations was developed for seven column regions.

In j addition, since the control rod pairs are located in the central

l. column and this column for a five column region is immediately l adjacent to the reflector, their reactivity worth is substantially l-l less than the other control rod pair groups. These groups are l typically worth less than 0.010 delta k.

Because of the low worth i

i and the analytical uncertainty, a larger range for the acceptance l criteria is required.

~

1 IE Fors St. Vrain 4 i

Technical Specifications

.L Amendment No. 77 Page 3/4 I-34

. l REACTIVITY CONTROL l'3/4.1.6 CONTROL ROD PAIR POSITION REQUIREMENTS-- SHUTOOWN l LIMITING CONDITION FOR OPERATION l

'n fueled regions shall be fully I 3.1.6 All control rod pairs i

l inserted and incapable of being withdrawn except:*

l 1.

Up to. two control rod pairs may be removed from fueled l

regions,-and E

' l 2.

Additional control rod pairs may be withdrawn for l

l

. SHUTDOWN MARGIN assessment or OPERABILITY tests, j APPLICABILITY: SHUTDOWN AND REFUELING l ACTION:

l A.

Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after each reactor shutdown, if more than l

two control rod pairs are not verified to be fully l

inserted, either:

l 1.

Insert at least all but two control rod pairs to the l

fully inserted position, or l

2.

Insert reserve shutdown material in at least those l

ragions where control rod pairs are not verified to

~

l be fully inserted, beyond the allowable two.

l B.

Subsequent to I hour after reactor shutdown, with less l

than the above requirements:

l 1.

Immediately suspend all operations involving-CORE l

ALTERATIONS, control rod pair movements resulting in l

positive reactivity changes or movement of l

IRRADIATED FUEL.

l l*

The SHUTDOWN MARGIN requirements of LCO 3.1.4 (for SHUTDOWN and 1

REFUELING) shall be maintained for all these control rod pair l

configurations.

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Vort[St,JVra"n d1 4

Technical Specifications 1

Amendment No. 77

,Page 3/4'1-35 V

U l SPECIFICATION. LCO 31.6 (Continued) l 2.

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

p l-a.

Insert any control rod pair capable of being l

inserted into fueled. regions and verify the I

SHUTDOWN MARGIN requirements of LCO 3.1.4 are l

met, or e

-l b.

Actuate sufficient reserve' shutdown material

'l into fueled regions to achieve the specified I

SHUTDOWN MARGIN.

l SURVEILLANCE REQUIREMENTS 1 4.1.6

= A.

Control rod pair positions for-all control rod pairs l

capable of being withdrawn shall be monitored for j'

compliance with LCO 3.1.6.A above, once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

- 7 l

B.

Following each reactor shutdown, each control rod pair

_l shall be verified to be at the fully inserted ' position l-by:

L l

1.

The rod-in position indication, or a

1 L'

-l 2.

The use of an independent control rod pair position l

verification method (e.g., watt-meter test).

l Control rod pairs known to be fully inserted prior to i

l the shutdown may be excluded. from

'the above l

l verifications.

l C.

Prior to the removal of more than one control rod drive

-l assembly,. the SHUTDOWN MARGIN shall be explicitly

.l calculated per the assumptions specified in SR 4.1.4.

3 l

0.

Upon full withdrawal of control rod pairs selected for l

removal, and prior to disabling their scram l

capabilities, the SHUTDOWN MARGIN shall be assessed byi l

1.

Withdrawing one or more additional control rod pairs l

with a calculated wortn greater tnan or equal to l

0.01 delta k, plus any calculated positive worth due I

to the temperature difference between the actual l

refueling temperature and 80 degrees F,

E.

- Cort St. Vrain #1

[F fi Technical Specificat-ions..

l'>'

+

~

. Amendment No. -77 Page 3/4 1-36 l

l-SPECIFICATION SR 4.1.6 (Continued)__

l.

2.

Verifying suberiticality, and'-

l 3.

Then reinserting the. additional control rod pairs.

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, Fort.St. Vrain 81 Technical: Speci ficat'lon s

=

a:

Amendment No. 77 j

Page 3/4 1-37 5

1 l-BASIS FOR SPECIFICATION L 0 3.1.6/SR 4.1.6 t

l This specification ensures that a sufficient number of control rod l pairs are. fully inserted into fueled regions to keep the reactor in a l shutdown condition (SHUTDOWN MARGIN greater than or equal to 0.01-l' delta k) in SHUTDOWN and REFUELING.

)

l Prior to refueling a region, the control rod pair in that region and l the control rod pair in the region next in sequence to be refueled I will.be withdrawn. Additional predesignated control rod pairs will l also be withdrawn. and suberiticality will be verified.

'The l calculated minimum reactivity worth of the additional predesignated I. control rod pairs is 0.01 delta-k plus the reactivity difference l between the new and spent fuel in the region to be refueled,.plus the l temperature defect between the refueling temperature and 80 degrees l F.

After.subtriticality has been verified, the predesignated control l rod pairs will be fully reinserted. Withdrawal of the predesignated l control rod pairs ensures a SHUTDOWN MARGIN of greater than or equal

-l to 0.01 delta k at 80 degrees F with new fuel loaded into the l refueled region.

This procedure will be followed until all control l rod pairs have been withdrawn without resulting in reactor l criticality, as determined with a calculated k(eff) not to exceed

-l 0.95 assuming all conditions specified in Specification 3/4.1.4.

l Making.all of the fully inserted co'ntrol rod pairs incapble of being l withdrawn ensures that an un-analyzed core configuration thich might l result in criticality will not exist.

In general, this is l accomplished by placing the reactor mode switch in the "off" position l or disabling the electrical supply to the motors. However, for any I specific control rod pair where analysis indicates inadvertent l withdrawal would result in criticality, this control rod pair is made j

l incapable of withdrawal by physically disconnecting its drive so that L

l its withdrawal is an incredible event for the purposes of analysis.

l This may be accomplished by lifting the power leads or other means l that involve more than just an administratively controlled clearance.

l I

L L

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' Fert.St. Vrain d' l

Technical' Specifications l4 ?

Amendment No.

77?

Page 3/4 1-38 l BASIS'FOR SPECIFICATI0tf LCO 3.1.6/SR 4.1.6 (Continued) l A~. SHUTDOWN-MARGIN of greater than or eoual to 0.01 delta k after l reactor shutdown (automatic scram or controlled) is ensured by the' 1

~

l hour ACTION to'either insert all but two control rod pairs or insert l reserve shutdown m'aterial in those regions where control-rod pairs l cannot be verified to be fully inserted, beyond the allowable two.

l Control rod pairs may be verified fully inserted by rod-in indication l (either rod-in limit, analog, or digital oosition indication), or by 4

4

.I other independent means (e.g., watt-meter test) as time permits. The 1 12 inch limit ensures reactivity control, as discussed in the Bases i for LCO 3.1.3; Experience has shown that a control rod pair which is -

l not fully inserted by a scram may still be fully inserted manually 1.with its control ved drive motor.

If the control rod pair cannot be l fully inserted with its drive motor, reserve shutdown material will l be inserted into that region, ensuring a SHUT 00WN MARGIN of greater l than.or equal to 0.01 delta k.

t r

l' If-any requirements of the LCO are not met while the reactor is in l SHUTDOWN or REFUELING, any control rod pair or fuel manipulations-l which would result in a positive reactivity addition wil.1 be

l. suspended immediately and within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> any withdrawn control rod i pairs.in fueled regions will be fully inserted.

If a SHUTDOWN MARGIN l of-greater than or equal to 0.01, delta k is not met-by fully l _ inserting the control rod pairs, sufficient reserve shutdown material l'will be inserted in fueled regions to achieve the specified SHUTDOWN i

l MARGIN.

This ACTION ensures a SHUTDOWN MARGIN of greater than or l equal to 0.01 delta k during reactor shutdown or refueling l operations.

l The reserve shutdown material provides an effective method of I

l_ reactivity control when. inserted into fueled core regions where the l control rod pairs have not been fully inserted. Because of the l_

l proximity to the control rod pairs, it has almost no additional worth 1:

l when inserted in regions where the control rod pairs are inserted.

l Therefore, to ensure an adequate SHUTDOWN MARGIN, it need only be I inserted into those fueled core regions where full insertion of the l control rod pairs cannot be demonstrated.

l Verifying control rod pair positions once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, ensures l that control rod pair position can be monitored during control rod l pair manipulations performed in refueling operations.

l After each shutdown, verifying that each control red pair is fully l

l inserted ensures that the position of each control rod pair is known l and that the SHUTDOWN MARGIN assessment is accurate.

Tech'nical Specifications' t

16 ' '

Amendment No.-

77.

Page 3/4 2-39 l BASIS _ FOR SPECIFICATION LCO 3.1.6/SR 4.1.6 (Continued)

F

-l Demonstrating that a SHUTDOWN MARGIN of greater than or equal to 0.01 l: delta k exists prior _to removing more than one control rod drive I assembly -from fueled regions ensures that criticality will not occur

-l and the SHUTOOWN MARGIN requirement's will be maintaired after the

~

l control. rod pair is removed.

i L

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Fort St. Vrain 01 c-Technical Specificat' ions P ge $ 414h 1

' l REACTIVITY CONTROL l 3/4.1.7 REACTIVITY CHANGE WITH TEMPERATURE i

l LIMITING CONDITION FOR OPERATION-4 l 3.1,7 The reactivity -change due to a CORE AVERAGE TEMPEkATURE l

increase between 220 degrees F and'1500 degrees F, shall be at l_

least as negative.as 0.031' delta k, but no more negative than l.

0,065 delta k throughout the REFUELING CYCLE.

1 APPLICABILITY:

POWER, LOW POWER, and STARTUP l-ACTION:

~With the reactivity change outside of the above limits, be i

in SHUT 00WN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, i

l. SURVEILLANCE REQUIREMENTS I 4.1.7 At the beginning of each REFUELING CYCLE the reactivity change-l-

as a function of CORE AVERAGE TEMPERATURE change (temperature l

coefficient) -shall be measured and integrated to verify that l

the measured reactivity change is within the above limits, a

l f

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Fort St. Vrain #1-

[

Technical Specifications L,x Amendment No. 77:

Page 3/4 1-41 s

i.

l. BASIS FOR SPECIFICATION LCO 3.1.7/SR'4.1'.7 y

l The negative temperature coefficient is an inherent safety mechanism

-t l that tends to limit power increases during temperature excursions.

I ~It is a stabilizing element in flux tilts or oscillations due, for l example, to xenon transients.

e l Fuel temperatures during a power excursion beginning from~a high l' power level are well within design limits regardless of tne magnitude l of the negative temperature coefficient, provided protective action l is initiated by a power level signal. However, if protective action j occurs much later, such as from a manual scram or actuation-of the.

l reserve shutdown system, peak fuel temperatures will be sensitive to l the magnitude of the negative temperature coefficient.

l Requiring a reactivity change at least as negative as 0.031 delta k l for a CORE AVERAGE TEMPERATURE increase from 220 degrees F to the l 1500 degree F temperatures asso:iated with the nominal RATED THERMAL o

l-POWER value, ensures temperature coefficients at least as negative as l those used in the FSAR accident analysis.

All control rod pair L'

l withdrawal transients assume a reactivity temperature defect of 0.028 l

l delta k which when, combined with an uncertainty of plus or minus 10%,

l yields the specified defect of 0.031 delta k.

I

-l The: maximum -reactivity temperature defeet of 0.065 delta k (0.072 l

l-delta k minus 0.007 delta k for uncertainty) assures that there is L

l sufficient reactivity control to ensure reactor SHUTDOWN in the

~ l unlikely event that all control roc' pairs cannot-be inserted and the I reserve shutdown system has been actuated.

l The reactivity worth of the reserve shutdown system was calculated to l be.0.130 delta k in the equilibrium core (FSAR Section 3.5.3).

From l' calculated excess reactivity data in Table 3.5-4 and Section 3.5.3 of l the FSAR it is seen that the maximum excess reactivity in the l equilibrium core with the CORE AVERAGE TEMPERATURE of 220 degrees F, l_Xe-135 decayed, Sm-149 built up, and 2 weeks Pa-233 decay, is 0.102 l delta k.

Assuming no control rods are inserted and the reserve l shutdown system has been activated, the excess SHUTDOWN MARGIN for l that excess reactivity is 0.028 delta k, (0.130 delta k minus 0.102-l delta k).

The calculated reactivity temperature defect for that l cycle is 0.044. delta k.

Therefore, if the reactivity temperature l defect were as large as 0.072 delta k (0.044 delta k plus 0.028 delta l k) reactor SHUTOOWN could be ensured for at least 2 weeks even for l the unlikely event that all control rods failed to inseTt, and the l reserve shutdown system was actuated.

L Fort St. V ain d!'

2 c.

Technical Specifications-L Amendment No.

77

1

~.-

Page 3/4 1-42

[. BASIS-FOR SPECIFICATION LCO 3.h7/SR 4.1.7 (Continued)

[

l.The major shifts in reactivity change as a function of core l temperature change will occur following refueling.

The specified l frequency of measurement ~ following each refueling will ensure that l the change of, reactivity a,s a function of changes in core temperature m

I will be measured on a timely basis,to evaluate the limit provided in

.i l LCO 3.1.5.

i.

l'The maximum value of reactivity temperature _ defect occurs at the

- l beginning of the cycle and slowly decreases through the cycle to a l minimum value at the end of the cycle.

Since.the measurement is made l at the beginning of a cycle and the minimum value occurs at the end-i of a cycle, a direct evaluation of the minimum reactivity temperature

l. defect cannot be made.

However, by comparing the calculated value-at I the beginning of the cycle with the measured value, an evaluation for l-compliance can be made using_the calculated value at the end of l cycle.

Performance of the Surveillance Requirement verifies the I assumptions used in the analysis.

L 1

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Fort St. Vrain #2 1

Technical Specifications j

Amendment No. 77 Page 3/4 1-43 s

l R_EACTIVITY CONTROI l 3/4.1.8 RESERVE SHUTDOWN SYSTEM - OPERATING ik l LIMITING CONDITION FOR OPERATION l l 3.1.8 All reserve shutdown (RSD) units shall be OPERABLE with:

i l

A.

At least 1500 psig pressure in their individual helium l

gas bottle supplies, and l

B.

At least 500 psig pressure in the Alternate Cooling l

ENethod (ACM) nitrogen bottles which provide a backup i

means of actuating the RSD hopper pressurization valves, j

l APPLICABILITY:

POWER,-LOW POWER, and STARTUP l ACTION:

l A.

With one RSD unit inoperable, operation may continue l

provided that an'0PERABLE spare RSD unit is available.

.l B.

With two or.more RSD units inoperable, restore all but-l one inoperable RSD unit to OPERABLE within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or l

be in at least 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 l

C.

The provisions of LCO 3.0.4 are not applicable for 1

changes between STARTUP, LOW POWER, and POWER.

Prior to l~

entry into STARTUP.from SHUTDOWN, all the requirements l..

I of this LCO must be met, without reliance on the

,~

l provisions contained in the ACTION statements.

l SURVEILLANCE REQUIREMENTS l

l l 4.1.8 The reserve shutdown system shall be demonstrated OPERABLE:

l A.

Once per 7 days by verifying that the pressure of each I

helium gas bottle is at least 1500 psig.

l B.

Once per 7 days by ve,rifying that the pressure of each l

ACM nitrogen bottle is at least 500 psig.

I

R J

,s of Fort St. Vrain

)

Technica~1 Specifications A-Amendment'No. 77 Page 3/4 1-44 lSPECIFICATIdNSR4.1.8(Continued)

I C.

Once per 92 days by:

l' 1.

Pressurizing each of the 37 RSD hoppers above' l

reactor pressure, as indicated by operation of the l

hopper pressure switch, I

2.

Operating the ACM quick disconnect couplings, and-l 3.

Performing-a CHANNEL ' FUNCTIONAL' TEST of 'the l

instrume.ntation which alarms at low pressure in the l-RSD actuating pressure lines.

1 D.

Once per 366 days by performing a CHANNEL CALIBRATION of I

the gas pressure instrumentation.

l E.

Following entry of condensed-water into any RSD system-I hopper (s) _(see LCO 3.1,1 ACTION H),

by performing the l

Surveillance Requirements identified in SR 4.1.9.E.

1 I

l i

l l

l l

1

}

1..

rort St. Vrain L

Technical Specifications Amendment No. 77 Page 3/4 1-45 lIBASIS.FOR SPECIFICATION LCO 3.1.8/SR 4.1.8 l The. reserve shutdown (RSD) system must be capable of achieving l reactor shutdown in the event that the control rod pairs fail to l insert.

j'After extended power operation, the RSD system'must add sufficient-l negative reactivity to overcome the. temperature defect between-1500 l and 220 degrees F, the decay of Xe-135, and some decay of Pa-233 to l U-233.

The buiidup of-Sm-149 also adds negative reactivity and is l taken into account in reactivity evaluations.

l The core reactivity increase due to core cooldown and Xe-135 decay l occurs within a few days and was calculated to be between 0.089 delta-lk and 0.081 delta k, at the beginning and end of the initial cycle, l respectively, and about 0.076 delta k for the mid cycle of the j equilibrium core. The reactivity increase is largest in the initial m

l core where the thorium loading is high and decreases through the l first' six cycles to a minimum value for the equilibrium core.

The I reactivity increase due to Sm-149 buildup and Pa-233 decay occurs l over several weeks to months and increases the. core excess reactivity l for the equilibrium core by about 0.007 delta k during the first 14

-l days, and by about 0.024 delta k after a few months, including full l Pa-233 decay.

Therefore, the reactivity control requirement for the l RSD system, including an allowance of 0.01 delta k for SHUTDOWN l MARGIN, -in the-absence of any* control rod pairs being inserted is l 0.098 delt k for the initial core and 0.093 delta k for the l equilibrium core after 14 days of Pa-233 decay and 0.121 delta k and l 0.110 delta k after full Pa-233 decay.

(FSAR Section :3.5).

l The calculated worth for the RSD system as noted,in FSAR Section l 3.5.3 is at least 0.14 delta k in the initial core, and 0.13 delta k l in the equilibrium core.

The worth of the RSD System with the I maximum worth RSD unit inoperable for those cases is at least 0.12 l delta k in the initial core and 0.11 delta k in the equilibrium core, I which is sufficient to ensure SHUTDOWN during the first 14 days of-i l Pa-233 decay.

l Generally, inoperable RSD units are capable of being restored to l-OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

However, in the unlikely event that I an inoperable RSD unit cannot be restored to OPERABLE within this l time, there is adequate time (at least 14 days due to the slow Pa-233 l decay as discussed in the BASIS for Specification 3/4.1.4) following l a shutdown using the RSD system, to allow for corrective action of l changing out a CR0 assembly.

A spare RSD unit is considered I available if.it is on site.

1 l

l

et M

Fort St. Vrait I

Technical Specifications

="

Amendment No. 77 Page 3/4 1-4o i

b l' BASIS FOR SPECIFICATION LC1 3.1.8/SR 4.1.8 (Continued) f

-l Two or.more RSD units may be inoperable for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to provide a l

l reasonable time for repair.- This is permissible because the control 1 rod pairs are available to shut down the reactor in the unlikely l event that a shutdown would be required during this short period of l time.

~j A minimum pressure of 1500 psig in the individual helium gas bottle l supplies is adequate because the pressure required to burst the l rupture discs is'1100 psig (FSAR Section 3.8.3).

The rupture discs 4

l are designed. and have been tested to burst at a differential I pressure of 165 plus or minus 50 psi.

t iA minimum pressure of 500 psig in the ACM nitrogen bottles is l adequate because the required set pressure is 220 psig.

A set I pressure of'220 psig is based on stroking a bank of 10 RSD valves one-l time and keeping the regulator fully open. This value also l compensates for minor line losses and system leakages.

I Each.of the 37 RSD hoppers shall be pressurized above reactor i pressure once per 92 days. Two redundant pressurizing valves will be l opened. using local test switches and the corresponding hopper l-pressure-observed to increase.

To prevent releasing absorber l material, the high pressure gas' cylinder is isolated and the l-pressurized actuating line is vented prior to the test.

L' l Pressurization is accomplished using test gas at a pressure i

i differential of approximately 40-70 psi above reactor pressure, which L

l 1s-below the 115 psi differential pressure required to rupture the l disc.

The hopper pressure should increase at least 10 psi above l reactor; pressure,'as indicated by the hopper high pressure alarm.

l

l. A CHANNEL FUNCTIONAL TEST will be performed on the low pressure alarm E

l instrumentation once per 92 days to ensure that the minimum required

l. rupture gas pressure can be monitored.

lA CHANNEL CALIBRATION will be performed on the gas pressure 1 instrumentation once per 366 days to ensure reliable monitoring of l the helium and nitrogen gas supplies.

I In the event that condensed water enters into any RSD system hoppers, l two RSD hoppers shall be functionally tested out of the core.

One

.l assembly will contain 20 weight percent and the other 40 weight l percent boronated material. The RSD hopper will be pressurized to the l point of rupturing the disc and releasing the poison material. The I material will be visually examined for boric acid crystallization and I chemically analyzed for boron carbide and leachable boron content.

a

.rcrt a;.:, rein na-

. Technical Specifications-Amendment No.

'77-

'+:

Page 3/4 1-47

- l REACTIVITY CONTROL l _ 3/4. I.9 RESERVE SHUTOOWN SYSTEM - SHUTDOWN l

l. LIMITING CONDITION FOR OPERATION

- l 3.1.9 Reserve shutdown (RSD) units in control i-od drive-assemblies i

for which control red pairs are capable of being withdrawn l

from fueled regions shall be OPERABLE (except RSD units in i

l-any control rod drive assemblies removed for l

refueling / repair) with:

l-A.

At least 1500 psig pressure in their individual-helium

' l gas bottle supplies, and l

B.

At least 500 psig pressure in the Alternate Cooling l.

Method (ACM) nitrogen bottles which provide a backup

= l means of actuating the RSD hopper pressurization valves.

l APPLICABILITY: SHUTDOWN and REFUELING l ACTION: With less than the above required RSD units OPERABLE, within l

24-hours either:

l A.

Return all control rod pairs in fueled regions (except I

the ones removed for refueling / repair) to the fully l

inserted position, or i

l-B.

Ensure SHUTDOWN MARGIN requirements are met (LCO 3.1.4),

l or l

C.

Insert sufficient RSD material to maintain SHUTDOWN l

MARGIN requirements.

G d

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tort St. Vrain #1 u-Technical Specifications.

Amendment No. 77 Page 3/4.1-48 m

~

t l SURVEILLANCE REQUIREMENTS l 4.1.9 The reserve shutdown system in fueled regions shall be l

-demonstrated OPERABLE:

-l

'A.

Prior to withdrawal and once per 7 days thereafter, for

.l those control rod pairs capable of being. withdrawn, by l

veri fying that the pressure of each required individual l

hopper helium gas bottle is at least 1500 psig.

l B.

Prior to withdrawal and once per 7 days thereafter, for l

those control rod pairs capable of being withdrawn, by I

veri fying-that the pressure of each required ACM l

nitrogen bottles is at least 500 psig.

l C.

Once per 366 days by performing a CHANNEL CALIBRATION of l

the gas pressure instrumentation.

l 0.

Once per 18 months by:

l 1.

Demonstrating that each subsystem is OPERABLE by l

actuating each group of pressurizing valves from the l

control room and verifying that the valves open.

l The capability of pressurizing the corresponding i

hoppers need not be demonstrated during this test.

l 2.

Performing a CHANNEL CALIBRATION of the RSD hopper l

pressure switches at the time of control rod drive l

preventive mainter.ance (SR 4.1.1).

l

'3.

Visually examining the pipe sections which eequire i

disassembly and reassembly within the refueling l

penetrations, after they have been disassembled for l

preventive maintenarce (SR 4.1.1),

and veri fying l

that there is no deformation or corrosion that could l

affect RSD system OPERABILITY.

l 4.

Functionally testing the RSD assembly most recently l

removed from the core.

The test consists of l

pressurizing the RSD hopper to the point of l

rupturing the disc and releasing the RSD material.

l The RSD material from the tested RSD hopper shall be l

visually examined for evidence of boric acid crystal l

formation and chemically analyzed for boron carbide l

and leachable boron content.

Failure of a RSO l

assembly to perform acceptably during functional l

testing or evidence of extensive boric acid crystal l

formation will be reported to the Commission within

-l 30 days per Specification 7.5.

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_ Fort St. Vrain di 1*

Technical Spe,cifications 3 #

Amendment No.

77-Page 3/4 1-49_

t l SPECIFICATION SR 4.1.9 (Continued) l l

E.

Following entry of condensed water'into any RSD system l_

hopper (s) (see LCO 3,1,1 ACTION H),

by performing the l

Surveillance Requirements identified in'SR 4.1.9.D 4 on l

two RSO assemblies, one containing 20 weight percent l'

boronated material-and the other containi.ng 40 weight l

percent boronated material. At the point in time when l

all of the regions having RSD hoppers with 40_ weight l

percent boronated material have been defueled, one,RSD 1

assembly containing 20 weight percent boronated material l

shall be tested.

I i

4 5

L 1

l I

Fo-t St. Vrain #1 Technical Specifications-4 Amendment No.

77 Page 3/4 1-50 l~BMIS FOR SPECIFICATION LCO 3.1.9/SR 4.1 9.

1 l The reserve shutdown (RSD) system must be capable of achieving l reactor shutdown in the event that the control rod pairs ' fail to l insert.

e l After extended power operation,-the RSD system must add sufficient l negative reactivity to overcome the temperature defect between 1500 l and 220 degrees F, the decay of Xe-135, and some decay of Pa-233 to l U-233.

The buildup of Sm-149 also adds negative reactivity and is I taken into account-in reactivity evaluations.

l The core reactivity increase due to core cooldown and Xe-135 decay l occurs within a few days and was calculated to be between 0.0S9 delta lk and 0.081 delta k, at the beginning and end of the initial cycle.

-l respectively, and about 0.076 delta k for the mid cycle of the

i. equilibrium core.

The reactivity increase is largest in the initial l core where the thorium loading is high and decreases through the l first six cycles to a minimum value for the equilibrium core. The l reactivity increase due to Sm-149 buildup and Pa-233 decay occurs l over several weeks to months and increases the core excess reactivity l for the equilibrium core by about 0.007 delta k during the first 14 l days, and by about 0.024 delta k after a few months, including full l Pa-233 decay.

Therefore, the reactivity control requirement for the l RSD system, including an allowance of 0.01 delta k for SHUTDOWN i

l MARGIN, in the absence of any control rod pairs being inserted is l 0.098 delta k for the initial core and 0.093 delta k for the

-l equilibrium core after 14 days of Pa-233 decay and 0.121 delta k and l 0.110 delta k after full Pa-233 decay.

(FSAR Section 3.5).

l The ca'lculated worth for the RSD system as noted in FSAR Section I-l 3.5.3 is at least 0.14 delta k in the initial core, and 0.13 delta k l

'l in the equilibrium core.

The worth of the RSD System with the l maximum worth RSD unit inoperable for those cases is at least 0.12

-l delta k in the initial core and 0.11 delta k in the equilibrium core, l which is sufficient to ensure SHUTDOWN during the first 14 days of ir L

l Pa-233 decay.

l-l l Generally, inoperable RSD units are capable of being restored to l OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

However, in the unlikely event that l

l an inoperable RSD unit cannot be restored to OPERABLE within this l time, there is adequate' time (at least 14 days due to the slow Pa-233 l

l decay as discussed in the BASIS for LCO 3.1.4) following a shutdown 1

l using the RSD system, to allow for corrective action of changing out l a CRD assembly.

A spare RSD unit is considered available if it is on i.

l site.

l l

l

(

Amendment No. 77 I

o Page 3/4 1-51 I BASIS FOR SPECIFICATION LCO 3.1.9/SR 4.1.9 (Continued)

I Ensuring SHUTDOWN MARG!N requirements for a CORE AVERAGE TEMPERATURE l greater than or equal to 220 degrees F is acceptable and provides j

l for changing out a Control Rod Drive (CRD) assembly, if necessary.

l Under normal conditions when the reactor has been operated for i several months (which is required for Pa-233 buildup), a CORE AVERAGE l TEMPERATURE greater than 220 degrees F is retained for a period of 2-l4 weeks even with the CORE AVERAGE INLET TEMPERATURE as low as 100 l degrees F.

This is adequate time for the replacement of a CRD l assembly.

l A minimum pressure of 1500 psig in the individual helium gas bottle j

l Supplies is adequate because the pressure recuired to burst the I rupture discs is 1100 psig (FSAR Section 3.8.3).

Tne runture discs l are designed and have been tested to burst at a differential l pressure of 165 plus or minus 50 psi.

lA minimum pressure of 500 psig in the ACM nitrogen bottles is l adequate because the reauired set pressure is 2;0 psig.

A set i pressure of 220 psig is based on stroking a bank of 10 RSD valves one I time and keeping the regulator fully open.

This value also l compensates for minor line losses and system leakages.

lA CHANNEL CALIBRATION will be performed on the gas pressure 1 instrumentation once per 366 days to ensure reliable monitoring of l the helium and nitrogen gas supplies.

l Once per 18 months, the most recently removed RSD hopper will be

} functionally tested and examined. The RSD hopper will be pressurized l to the point of rupturing the disc and releasing the poison material.

l The material will be visually examined for boric acid crystallization l and chemically analyzed for boron carbide and leachable boron l content.

l LCO 3.1.9 only requires RSD units to be OPERABLE for those control I rod pairs capable of being withdrawn because tne worth of the control I rod pai r( s) removed from the PCRV has been accounted for in the l SHUTOOWN MARGIN and the worth of the RSD material in regions whose l control rod pair are inserted adds little to the SHUTDOWN MARGIN.

2 l The ACTION time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is adequate because the reactor has l already been shutdown and the SHUTDOWN MARGIN requirements met, 1 versus verifying SHUTDOWN MARGIN requirements immediately after a l

l shutdown.

l Once per 18 months, each group of pressurizing valves will be I actuated from the control room to verify that the valves open.

m k

Fert St Vredn #;

Tectnical Specifications o

Amendment No. 77 Page 3/4 1-52 i BASIS FOR SPECIFICATION LCO 3.3.9/SR 4.1.9 (Continued)

_..m l Once per 18 months, the RSD hopper pressure switches which measure I the pressure differentiel between the hoppers and the reactor will be l calibrated as individual control and orifice assemblies are removed l from the reactor for servicing and maintenance.

These switches alarm j high pressure for pressurization testing or actual system operation.

l In the event that condensed water enters into any RSD system hoppers, I two RSD hoppers shall be functionally tested out of the core.

One l assembly will contain 20 weight percent and the other 40 weight I percent boronated material.

The RSO hopper will be pressurized to l the point of rupturing the disc and releasing the poison material.

l The material will be visually examined for boric acid crystallization l and chemically analyzed for boron carbide and leachable boron l content.

I If there is no RSO assembly containing 40 weight percent boronated l material over a fueled region, its negative reactivity is not l insertable, and as such, testing it would be meaningless, and testing l one assembly containing 20 weight percent boronated material is l sufficient.

l The refueling penetration pipe sections will be visually examined for i deformation and corrosion following disassembly for refueling or I maintenance.