ML20101U041

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Control Rod Sys Operability Evaluation
ML20101U041
Person / Time
Site: Fort Saint Vrain Xcel Energy icon.png
Issue date: 01/31/1985
From:
PUBLIC SERVICE CO. OF COLORADO
To:
Shared Package
ML20101U035 List:
References
NUDOCS 8502060544
Download: ML20101U041 (35)
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V CONTROL ROD SYSTEM OPERABILITY EVALUATION TABLE OF CONTENTS ABSTRACT INTRODUCTION I.

FSAR REVIEW................................................

2 A.

SYSTEM DESCRIPTION....................................

2 B.

DESIGN BASES..........................................

2 Control Rod......................................

2 Drive Mechanism..................................

4 Rod Control And Position Indication..............

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C.

ESSENTIAL SAFETY FUNCTIONS AND COMPONENTS............ 8 D.

CONCLUSIONS.......................................... 13 II.

TECHNICAL SPECIFICATION REVIEW............................ 15 A.

LCO, SR 0VERVIEW...................................... 15 B.

CONCLUSIONS AND RECOMMENDATIONS...................... 19 III.

CONTR3LLED DOCUMENTS REVIEW................................

21 A.

SAFETY RELATED LIST.................................. 21 B.

CONCLUSIONS.......................................... 23 - FSAR Scram Parameters - Technical Specification Scram Instrumentation Requirements - Technical Specification Scram Surveillance Requirements F

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CONTROL ROD SYSTEM OPERABILITY EVALUATION

= ABSTRACT:

This report summarizes the results of a detailed review and evaluation of the existing licensing basis for the reliability of the Fort St. Vrain Control Rod System. All FSAR design and safety considerations were reviewed to identify the significant design bases, and the essential safety functions and components required for accident analyses.

Once identified, these functions and components were evaluated for consistency with Technical Specification requirements, controlled documents and procedures, and plant operational experience.

In general, the licensing basis was found to be consistent with Technical Specification requirements and design documents.

However, specific
changes, additions, and evaluations are recommended in maintaining the original level of reliability in view of plant operational experience and continuing engineering investigations.

INTRODUCTION:

This-repert is organized to highlight the three main areas reviewed in evaluating the licensing basis for control rod reliability:

FSAR REVIEW, TECHNICAL SPECIFICATION REVIEW, and CONTROLLED DOCUMENTS REVIEW.

Specific conclusions and recommendations are included following the review of each area.

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FSAR REVIEW A.

SYSTEM DESCRIPTION The control rod system functions to control and safeguard the fission process occurring in the reactor.

The main components of the centrol rod system consist of the control rod, the drive mechanism, and the control and position indication circuitry.

The control rod consists of eleven (11) baron carbide cannisters and a tube type shock absorber attached along a metal spine suspended in the core from steel cables.

The design considerations are specifically described in FSAR Section 3.8.1.2.

The important design considerations are related to boron content, ruggedness of design, and component design life assumptions.

-The drive mechanism primarily consists of the drive motor, motor break, reduction gearing and bearings, guide

pulleys, cable
drum, limit switch cams, position potentiometers, guide tubes, and a velocity limiting three phase capacitor array.

These components are discussed in FSAR Section 3.8.1.1.

The drive mechanism is designed to be fail-safe under all postulated accident and operating conditions, allowing for free-fall gravity insertion at all times.

The rod control and position indication system consists of the automatic and manual

controls, associated circuitry, interlocks, power sources, sensors, and various relays, which provide for normal reactivity control and indication as well as abnormal reactor protective actions.

Reactivity control is described in.

FSAR Section 7.2.2, and protective actions in Section 7.1.2.

The automatic and manual scram capabilities are considered essential.

B.

DESIGN BASES The primary FSAR design bases and major assumptions for ensuring the reliability of the control rod system have been identified as listed. The design bases which are considered essential for performance of the scram safety function, as identified by accident analyses, have been identified by an asterisk (*).

Control Rod i

  • 1.

Individual boron loadings are 0.48 gm/cm' for the 8

inner nineteen (19) and 0.63 g/cm for the outer eighteen (18) rod

pairs, 30 and 40 wt. %

respectively (3.8.1.2, 3.5.3.1).

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  • 2.

The overall control rod worth and configuration, considering fuel and poison loadings, must be able to ensure suberiticality, with a minimum shutdown margin of 0.01 AK, under all conditions with the maximum worth rod pair withdrawn (3.2.2.3, 3.5.3.1, 3.2.3.2).

(See Technical Specification LCO 4.1.2.)

  • 3.

The structural integrity, flexibility, and overall dimensions will be maintained while exposed to the normal reactor operating environment, such that satisfactory operation, helium flow,- and free fall insertion are sustained (3.2.2.6, 3.8.1.2).

  • 4.

The normal operating environment for the-control rod will not exceed 1300*F (3.8.1.2) or 10vpm total oxidant impurities (CO, CO2, - H2O) during normal continuous operation (A.9.2.1, 4.2.1, 3.2.3.3, i -

3.2.3.5).

(See LCO 4.2.10,4.2.11)

  • 5.

The crushable tube-type shock absorber is designed to absorb the energy of a falling control -rod, due

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to cable or spine failure, such that the integrity of the boron cannisters and bottom reflector element is maintained (3.2.2.6, 3.8.1.2).

6.

The design life of the control rod is six (6) cycles (1800 effective full power days (EFPD)) of full power operation (3.8.1.2).

  • 7.

The maximum rod pair worth in the event of an accidental rod pair withdrawal,

'during all anticipated configurations, will.

result in a transient less severe than the reactivity accidents evaluated in Section 14.2 (3.5.3.1).

  • 8.

Under the design environmental conditions, the clearances, low drag

forces, and dry-film lubrication make the probability of galling or binding of the cables in the guides extremely unlikely (3.8.2).

9.

Cable fatigue life calculations show a life of approximately 1 x 10' jogs.

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Drive Mechanism

  • 1.

The coqtrol rod drive mechanism provides for free-fall rod insertion under loss of AC motor or DC brake power conditions (3.2.2.6, 3.8.1.1).

2.

The CRD motor rotates under -the influence of a capacitor array to limit the speed 'of control rod insertion during gravity driven scram conditions (3.2.2.1,3.8.1.1).

3.

Environmental operating conditions are maintained within acceptable limits based on design thermal barriers, radiation shielding, and normal operation of the penetration purge flow and liner cooling systems (3.8.1.1).

4.

Radiation shielding and, primary coolant activity levels are designed to limit drive mechanism radiation levels to 1 rad /hr under normal continuous operating conditions (3.8.1.1.1).

  • 5.

The maximum temperature

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mechanism which might inhibit the scram function is 272 F.

6.

The normal penetration purge flow is designed to be approximately 5 lb/hr/ penetration.

7.

Presence of foreign particles and debris, both metallic and molybdenum disulfide, was observed during the original prototype testing. However, it was specifically evaluated and determined to have no significant effect on drive performance based on design provisions which limit ingress and accumulation.

  • 8.

All bearing and gear materials, fabrication, and special dry film lubrication have been proven through extensive testing to maintain satisfactory operation in the purified helium environment (3.8.1.1.1).

  • 9.

Gravity free fall capability is based on an initiating load of 120 lbs. per cable (Page 3.8-5).

  • 10. The drive mechanism is designed to withstand the maximum seismic dis;urbance or Design Basis Earthquake without loss of function (3.8.2).
  • 11. The total scram insertion tlme is approximately 152 seconds (3.5.3.1). [

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  • 12. The maximum reactivity insertion rate is about 0.001 AK/ft, based on a normal complete rod pair withdrawal time of approximately 180 seconds (3.5.3.1,3.6.7).
13. Operation of the-control rods by the control rod drive system, including representative numbers for scram operations, is possible for at least the six cycle (1800 EFPD) minimum life of the control rods

-(3.2.2.6).

14. The prototype testing, initiated to ascertain the reliability of the control rod system, simulated the expected long term operating conditions of temperature and helium, with less than 10 VPM oxidant impurities, and no radiation effects. In the shim mode, the prototype demonstrated some 200 years of service life or 33 times its expected service life (6 years) (A.9.2.2).
15. The rod drives were to receive inspection and refurbishment as necessary (A.9.2.2).

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Rod Control And Position Indication

  • 1.

A rod withdrawal sequence interlock prevents rods from being withdrawn out of sequence at power levels between 1 and 5% rated power (3.5.3.1, 7.1.2.2, 7.2.2.1).

(See Technical Specification LCO 4.1.3) 2.

The control and position indication system is utilized to establish and measure the core power level (7.2).

  • 3.

Partial control rod insertion is required to prevent endangering fuel particle integrity for region peaking factors greater than 1.83(3.2.3.1). (See

. Technical Specification LCO 4.1.3) 4.

The runback controller is allowed to insert rods only(7.2.1.2).

5.

Rod control actuator switch interlocks and power supply load sensors ensure that not more than one rod-pair may be moved simultaneously outward (7.2.2.1,7.2.2.3).

6.

Each of the thirty-seven (37) control rod drives is equipped with two (2) potentiometer type position transmitters, one providing continuous analog indication for each rod and one providing digital indication on a selective basis in the control room (7.2.2.3,7.2.2.1).

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7.

In addition to Item 6 (above), each rod pair is equipped with three pairs of limit switches which provide control room indication of individual full in/ full out position, outward /inward rod motion, and slackcable(7.2.2.1,7.2.2.2,3.2.2.6,C.13).

  • 8.

Means must be included in the control room to monitor and control the reactivity status of the reactor (7.2.2.1, C.13.1).

(See Technical Specification LCO 4.1.8) 9.

Excessive deviation between rod pairs in a group is alarmed for rod deviations greater than 2 1 1 ft (Page 3.6-19, Page 7.2-9 and Section 7.2.2.1).

  • 10. To prevent undesirable flux and temperature distributions, partial rod insertion, with the exception of the regulating rod pair, shall be limited to two groups at any position (separated by at least 10 ft), six pairs up to 2 ft, and the two runback groups (six pairs) at any position not to exceed 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (3.2.3.1).

(See Technical Specification LCO 4.1.4).

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  • 11. The automatic scram circuitry provides three independent sensing circuits for each scram parameter, and is based on a general 2 of 3 logic system up to the final trip logic (7.1.2.1).
  • 12. Direct DC brake power supply interruption is provided through Manual Scram capability, independent of the automatic system (7.1.2.1).
13. Relays in the rod brake circuitry deenergize contactors in the rod motor circuit to ensure scram functions (7.1.2.1).
14. Manual push-button bypass circuitry is provided to allow powered insertion of a bound rod following a scram (7.1.2.1).

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  • 15. Remote manual scram capability is provided in the switchgear room to effect plant shutdown in the event the control room becomes uninhabitable (7.1.2.3).
  • 16. The reactor mode switch (RMS) is provided as a backup to manual scram (7.1.2.3).

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  • 17. The automatic scram parameters are defined as shown in Attachment I to this report (Table 7.1-2).

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ESSENTIAL SAFETY FUNCTIONS AND COMPONENTS 1

Through review of the FSAR accident analyses, with respect to the previously listed design bases, essential safety functions j

and components ~ have been identified along with general-L conclusions regarding accident evaluations.

i l-Environmental Disturbances iI (Sections 14.1, 1.4, and 10.3) i

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Of-all the FSAR accident evaluations, the environmental jl disturbance accidents are probably the most significant in terms of the impact on equipment requirements. All plant structures, systems, and components.have been divided into two groups, Class I and Class II, based on their importance to

. safety during environmental accidents. Of the environmental accidents,.the Design Basis Earthquake and Maximum Tornado were considered limiting, i

Class I equipment was specifically defined through evaluation of an encompassing accident involving a

Design Basis 1

Earthquake or Maximum Tornado, which-are evaluated to. include i

the failure or loss of: outside electric power, main turbine, deaerator, all three boiler feedpumps, all condensate pumps, auxiliary boiler and backup auxiliary boiler feed pumps, main condenser, main and service water cooling towers, and pipin; i

and: equipment downstream of the main steam bypass valves.

Under these conditions, items whose failure or damage could have resulted in:

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1) Release of abnormal quantities of radioactivity,
11) Interference with' safe reactor shutdown, or r

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were designated Class I.

The Class I list included certain considerations for redundancy, accident mitigation, and single i

failures where considered appropriate (10.3.10). The minimum requirements for cooldown of the plant under these~ conditions have been defined by another list of equipment items termed 1

Safe Shutdown, which is a subset of the Class I List (10.3.9 and 14.4.2).

Thus, all Class I items, with the exception of the fuel handling machine, are designed to withstand both the Design Basis Earthquake and the Maximum Tornado without unsafe damage or loss of safety function. (See Design Documents SR i

L 6-1 and SR 6-2) i All other plant structures, systems, and components were j

designated Class.II.

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7 The " control and orificing a'.semblies" are considered Class I and the " control rod drives" are considered required for safe shutdown cooling, as designated in Table 1.4-1 and 1.4-2 2

respectively. The ability of the control rod to drop freely into the core under worst case core misalignment conditions following an earthquake, is specifically evaluated in Section 14.1.1.

The conclusions of Section 14.4.2 regarding acceptable safe shutdown cooling, assume that a scram is achieved immediately following the event. This is consistent with Section 7.3.9, which requires immediate reactor shutdown following seismic instrument indication that a disturbance of the magnitude of the Design Basis Earthquake, 0.10g, has occurred.

The critical safety functions, for these conditions, would be those responsible for the scram functions. Scram, under the postulated conditions, can be assumed to occur automatically or manually within ten minutes after the event, as evaluated in Sections 10.3.3 and 10.3.1, respectively. (See Technical Specification LCO 4.4.1)

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i Reactivity Accidents (Section 14.2)

The FSAR evaluated reactivity accidents initiated by any of the

.following conditions:

1.

Excessive removal of control poison, 2.

Loss of fission product poisons,.

3.

Rearrangement of core components, 4.

Introduction of steam into the core, and 5.

Sudden decrease in reactor temperature.

From these. evaluations, it is concluded that the accidental withdrawal of control poison results.in the worst reactivity accidents.

Ten specific protective actions or lines of defense against the rod withdrawal accidents are provided, of which nine are considered effective during a startup accident, and five effective during power operation.

The inherent protective design

. features considered are the maximum reactivity addition rate of 0.00009 AK/sec., and the available scram reactivity, which is always sufficient to achieve subcriticality with a 0.01 AK shutdown margin with due regard for inoperable rod pairs. (See Technical Specification LCO 4.1.2) 10 -

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a Three main rod withdrawal accideniis are specifically reviewed:

1). Maximum Worth Control Rod Pair Withdrawal at Full Power,

11) Maximum Worth Control Rod Pair Withdrawal at Source Power,

.and 111) Simultaneous: Withdrawal of All Thirty-Seven Rod Pairs.

1) -The power range accident assumed three sequential

. lines of defense: automatic scram at 140% rated power as initiated by-the. power range channels, manual scram.after 60 seconds, and hot reheat-steam temperature automatic scram at 1075*F after 105 seconds.

Only when protective action-is not initiated prior to the-1075'F reheat steam temperature limit is fuel failure assumed to occur.

However, it is concluded that the 2% fuel particle failure would result in less than design primary coolant activity levels, and core. shutdown /cooldown and PCRV integrity would not be impaired.

(See Technical Specification LCO 4.4.1)

11) Assuming the sequential failure of four specified lines of defense,the source power accident was assumed terminated by a scram at 140% rated power.

The consequences of a 0.047 AK source power insertion was evaluated with no fuel particle failure expected. (See Technical Specification LCO 4.1.3) 111) The simultaneous rod pair withdrawal accident (37) was considered incredible due to the specific protective design features including control rod acutator switch interlocks, and rod motor power supply line load sensors.

For.the ' limiting conditions of 0.0029 Ak/sec. reactivity insertion, 180 second total withdrawal time, and 150 second rod insertion time, a scram initiated at 140% power will-not lead to fuel failure nor any other condition endangering the safety of the plant. (See Technical-Specification Surveillance Procedures SR 5.1.la-A/5.4.1.4.4.b-R-Load Sensor, Scram and Withdrawal Rate, and 5.4.1.4.4.a-P.-Hand Switch Interlocks) l I

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L Design Basis Accidents j

(Sections 14.10 and 14.11)

For Design Basis. Accident No. 1, permanent Loss Of Forced Circulation, the FSAR assumes an automatic scram on "two loop trouble" occurs upon initiation of the event.

Following scram, the core fission product afterheat is expected to result in peak temperatures of 2980*C.for the center of the.

active core. The boron compact loadings of 30 and 40 wt for inner and outer rods, were specifically evaluated and determined to maintain the structural integrity of the boron compacts thus ensuring that no major loss of poison material would occur (0.3.3).

The analysis of Design Basis Accident No. 2, Rapid Depressurization, assumes that automatic scram is initiated by the load programed PCRV pressure - Low, 50 psig below normal at 650 psig from full load. Neither.the event initiation nor conditions.following the event are considered to impair the reactor shutdown systems, control rods and reserve shutdown material.

(See Technical Specification LCO 4.4.1 Scram Parameters snd Settings)

Steam Leak Accidents (Section 14.5)

For the various limiting steam generator leaks analyzed in the FSAR, automatic scram is assumed to occur following correct operation of any one of three safeguards: high moisture (2 inputs), high pressure, or manual steam generator dump and i

scram.

These scram parameters are assumed to be operable to initiate corrective action within approximately 100 seconds following the event. (See Technical Specification LC0 4.4.1)

Other Accidents Other abnormal conditions such as loss of purge flow (14.6.1.1), cable failure (3.8), and loss of power have been evaluated and determined not to impair the shutdown function of the control rod system.

In the incredible event of total inoperability of the control rod system, the reserve shutdown system is adequate and independently redundant to achieve shutdown conditions from anyoperatingcondition(3.8.3).

(See Technical Specification LCO4.1.6andSR5.1.2) i L

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CONCLUSIONS-1.

AlthoughJthe control rod system was adequately evaluated to remain fail-safe under loss of purge flow conditions, i

purge flow was a design consideration for normal, continuous power operation for minimizing the effects of primary coolant. in the CRD motor area. Therefore, the proposed orifice motor plate and window seals will be installed to reduce purge flow requirements.

2.

Due to the concerns regarding control rod temperature, control rod temperature will be monitored on a regular i

basis.

3.

The control rod cable failure and corrective actions should be evaluated for impact on FSAR design life and operating environment assumptions. A 10CFR50.59 Safety Evaluation has been written for changeout of the material.

4.

The FSAR specifically considered both the ability to differentiate between rod motor withdrawal and insertion characteristics, and the ability to identify bound rods j

by measuring rod motor characteristics.

The proposed watt-meter and Back-EMF testing capabilities are being evaluated and formalized for use in predictive / preventive maintenance programs.

5.

Control Rod Drive refurbishment efforts have specifically i

identified as left acceptance criteria for ' design considerations related' to position indication, primary and secondary penetration seal leakage and scram time.

t 6.

CRDOA serial numbers will be verified and tracked to assure inner and outer ring boron loadings are maintained in accordance with the DBA-1 analysis.

7.

Recent investigations have determined that the major consideration in the observed failures to scran was long term control rod drive degradation. From FSAR design life considerations, the control rod absorber section was considered the limiting factor. The control ~ rod shock absorber was later defined as the limiting component of the control rod, due to neutron embrittlement. Once the design life of the control rod shock absorber was identified (1800EFPD), the drive mechanism was then prototype tested for performance over this expected service life.

However, actual operating experience has shown that normal degradation of the drive occurs t

independently of EFPD accumulation.

Periodic CR00A 1

performance monitoring will' be implemented to provide l

adequate information to detect significant degradation.

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Since all limiting accident analyses assume that automatic or manual scram is initiated early ~in the accident, performance degradation type failure would not need to be addressed provided that periodic testing and preventive / predictive maintenance programs are implemented.

Therefore, accident reanalysis is not necessary.

9.

The reserve shutdown system was designed to provide an alternate, independent means of shutting the reactor down from any operating condition without movement of the control rods. To ensure this capability, examination of reserve shutdown material will be included as a part of the CRDOA preventive maintenance program to verify that material bridging or agglomeration is not occurring.

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TECHNICAL SPECIFICATION REVIEW A.

LCO. SR OVERVIEW The Technical Specification requirements and corresponding procedures related to the control rods and the reserve shutdown system were reviewed to ensure that the identified FSAR analyses limits are incorporated, that the existing limits are consistent with FSAR analyses, that LCO's have appropriate SR requirements, and that SR requirements are maintained through appropriate SR procedures.

The LCO, SR, and SR procedure matrix was identified as follows:

LCO 4.1.2 Operable Control Rods SR 5.1.1 Control Rod Drives Surveillance SR 5.1.la-A/

Control Rod Scram Test / Multiple 5.4.1.4.4.b-R Rod Pair Withdrawal Check SR 5.1.lb-M Control Rod Operability SR 5.1.4-W-P Core Reactivity Status Check LCO 4.1.3 Rod Sequence SR 5.1.5 Withdrawn Rod Reactivity Surveillance SR 5.1.5-RX Control Rod Reactivity Worth LCO 4.1.4 Partially Inserted Rods LCO 4.1.8 Reactivity Status SR 5.1.4 Reactivity Status SR 5.1.4-W-P Core Reactivity Status Check LCO 4.4.1 Plant Protective System Instrumentation See Attachment 2 SR 5.4.1 Reactor Protective System See Attachment 3 SR 5.4.1.1.1.a-RP Manual (Control Room)

Scram Test SR 5.4.1.1.2.a-MP Manual (I-49)ScramTest.

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Startup Channel Scram Test 5.4.1.4.1.b-P SR 5.4.1.1.3.c-R Startup Channel Scram Calibration

.SR 5.4.1.1.4.b-M/

Linear Power Chanr.e1 5.4.1.4.2.b-M Scram Test.

SR 5.4.1.1.4.c-D/-

Linear Power Channel 5.4.1.4.2.c-D Heat Balance Calibration SR 5.4.1.1.4.d-R/

Linear Power Range 5.4.1.4.2.d-R Channel Calibration SR 5.4.1.1.5.b-P/

Wide Range Power 5.4.1.4.3.b-P Channel Test SR 5.4.1.1.5.c-M/

Wide Range Channel 5.4.1.4.3.c-M Heat Balance Calibration SR 5.4.1.1.5.d-R/

Wide Range Power 5.4.1.4.3.d-R Channel Calibration SR 5.4.1.1.6.c-R Primary Coolant Moisture Scram Calibration SR 5.4.1.1.6.e-M Primary Coolant Moisture Instrumentation Sample Flow 1

Alarm Functional Test SR 5.4.1.1.7.a-M Primary Coolant Moisture Scram Test SR 5.4.1.1.8.b-M Reheat Steam Temperature Scram Test SR 5.4.1.1.8.c-R Reheat Steam Temperature i

Scram Calibration SR 5.4.1.1.9.b-M/

Primary Coolant Pressure 5.4.1.2.9.a-M Scram Test SR 5.4.1.1.9.c-R Primary Coolant Pressure Scram Calibration i

SR 5.4.1.1.10.b-M Circulator Inlet Temp.

Scram Test i

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I* .y r 5 SR 5.4.1.1.10.c-R Circulator Inlet Temp. i Scram Calibration 1 .SR 5.4.1.1.11.a-M Hot Reheat Header . Pressure Scram Test SR-5.4.1.1.11.b-R Hot Reheat' Header Pressure Scram Calibration SR 5.4.1.1.12.a-M'. Main Steam Pressure Scram Test SR 5.4.1.1.12.b-R Main Steam Pressure Scram Calibration SR 5.4.1.1.13.a-M Two Loop Trouble Scram Test t SR 5.4.1.1.13.b-R Two Loop Trouble Scram-Test SR 5.4.1.1.14.a-M Plant 480V Power Loss Scram Test SR 5.4.1.1.15.b-M High Reactor Building Temperature (PipeCavity) Scram Test SR 5.4.1.1.15.c-R High Reactor Building Temperature (Pipe Cavity) Scram Calibration LCO 4.1.6 Reserve Shutdown System SR 5.1.2 Reserve Shutdown System-SR 5.1.2ad-Q Reserve Shutdown Hopper Pressure Test SR 5.1.2a-W ACM Nitrogen Backup Bottle Pressure i SR 5.1.2bd-A Reserve Shutdown Hopper low Pressure Calibration SR 5.1.2c-X Reserve Shutdown Assembly ' Functional Test SR 5.1.2e-X Reserve Shutdown Hopper Pressure Switch Calibration SR 5.1.2f-X Refueling Penetration Examination SR 5.1.2g-R Reserve Shutdown Valve Operability Test 17 - n .a., -..~.m-.- --mn,w.

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~. LCO's 4.1.2 and 4.4.1 ensure that the available scram reactivity worth and automatic / manual initiating actions respectively, are maintained functional in accordance .with the accident analyses of the FSAR. The scram parameters of LCO 4.4.1 and associated surveillance requirements are attached. LCO's 4.1.3 and 4.1.4 define design startup and power operation requirements which must be verified to ensure safe power assention and continuous power operation. LCO.4.1.4 is controlled administrative 1y and thus does not have a specific surveillance requirement. .: + ~. y..m c r. .w. v-~~ q B. CONCLUSIONS AND RECOMMENDATIONS The recommendations made below will be evaluated as a part of the Technical Specification Upgrade Program. 1) LCO 4.1.2 basically. requires that control rods be " operable" or " fully inserted" to verify available shutdown margin. Although the LCO states that these conditions must be met during power operation, the basis and the FSAR clearly require that they be met at all times. Therefore a change to the applicability of the LCO is recommended to make it consistent with the FSAR. The allowable actions in LCO 4.1.2 when withdrawn and partially inserted control rods are determined to be inoperable should be stated, along with the requirement to verify compliance within a certain period follow!ng rod inoperability. Per the basis of LCO 4.1.2, a control rod is considered operable if it demonstrates scram capability or is fully inserted. 2) SR 5.1.1 should be revised to adequately address the determination of scram capability for both withdrawn rods and partially inserted rods and position verification of fully inserted rods. Control rod position indication is also necessary to verify compliance with LCO 4.1.4, LCO 4.1.8, and the basis for LCO 4.1.2. It is therefore recommended that indication discrepancies and requirements be specified in SR 5.1.1 as well. 3) Provisions for acceptable alternate scram capability testing and rod-in position verification testing should be added to the Technical Specifications. 4) Provisions should also be added to include periodic checks of a representative sample of the control rod drive temperature indicators to ensure that the maximum temperature rating of 272"F is not exceeded during power operation. 5) The criteria defined in the basis for LCO 4.1.3 are actually design safety requirements and should be contained in the Specification section so that it is clear that these limits are not to be exceeded. 6) SR 5.1.5 for the measurement of control rod worths during cycle startup, should clearly state that a comparison of measured and predicted rod worths is required and that a 20% discrepancy is acceptable as specified in the procedure. The *20% acceptance criteria should be explained in the bases., u.

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7)' LCO 4.1. 4' is controlled administrative 1y and thus does not have an applicable surveillace requirement.- This LC0 should specify appropriate actions for exceeding limits .and allow specific time periods for achieving compliance. 8) All FSAR scram parameters are adequately ccntrolled and tested'per LCO 4.4.1 and SR_5.4.1. 9) The . reserve shutdown. system LCO, SR, corresponding procedures, and anticipated corrective actions are considered adequate to demonstrate and ensure the operability of the system. However, recent problems with system s' ggest the need for periodic examination of this u the material to monitor and detect long term degradation. Technical Specifications should be developed to require that one low and one high boron content hopper be - functionally tested on a refueling cycle basis and that the material collected undergo visual and chemical examination. 4 9 I a e i 0

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,.e,;- 9-2, s-III.. CONTROLLED DOCUMENTS REVIEW A. SAFETY RELATED LIST Control Rod Assembly The control and orificing assembly, as specified in FSAR 1 Table-1.4-1, is equipment ites D-1201 on -the Safety i Related Equipment List (see Dwg. D-1201-940). The Safety -l Related List includes all components which have been designated Class I. This assembly is designated seismic type 2, and environmental I.D. 5,' meaning that the item must function only following a seismic event, and that it is' required for safe shutdown (Dwg. 01200-100). Drive Mechanism The control rod drive e.echanics is not separately listed in.the Safety Related List, even though it is specifically listed as safe shutdown in FSAR Table 1.4-2. This is due to the fact that the whole assembly is listed as-Class I, Safe Shutdown. However,=for FSAR purposes, it is clear that the only part of the assembly required to remain operable for scram capability and Safe Shutdown, is the drive train assembly. The control rod absorber sections and power supplies are considered fail-safe. (See Surveillance Procedure SR 5.1.16-M) The rod motors are powerad directly from the Control Rod Drive Motor Control Centers 1 and 2 (N-9225, N-9226), through Reactor MCC's 1 and 3, (N-9229A, N-9231) which are all on the safety related list. i e l i i I l

t Scram Circuitry The protective Scram Circuitry is based on hindrance logic; the protective action is caused by loss of signal. The control rod brake power supply from Instrument Buses 1 and 2 (N-9237, N-9238), is normally supplying power to the control rod brakes, and can be interrupted by one of the following actions: 1. A scram signal from the PPS circuitry grounds out or de-energizes control power to the relay coils in the brake power supply lines, which causes the contacts to open, disconnecting the brake power supply and releasing the brake mechanism (Dwgs. IB-93-6 and D169-2951). The PPS contacts in the brake power supply are XM93125-1, -2, XM93126-1, -2, and XM93127-1, -2. The manufacturer is Square D, Model #CL7002-TG-2. They are listed on the safety related subtier component list as Subt-313, seismic type 1 (function both during and following a seismic event), environmental I.D.-3 (required for safe shutdown-located in three room control complex). (See Technical Specification Surveillance Requirement SR 5.4.1) 2. Numerous manual scrams may be initiated as a backup to the automatic scrams. The three predominant methods for manual scram are: actuation of the manual scram switch, HS-9330 on I-03; positioning the Reactor Mode Switch to 0FF, HS-1216 on I-03; and depressing 2 of 3 pushbuttons in the switchgear room, HS93372, HS93373, and HS93374, on I-49. These hand switches are all on the safety related list and are classified as seismic type 1, environmental I.D.-6 (Class I but not required for safe shutdown cooling-environmental qualification required for loss of air conditioning). (See Surveillance Procedure SR 5.4.1.1.la-RP and SR 5.4.1.1.2a-MP) In addition to de-energizing the brake circuit, the brake power supply also supplies control power to a set of relays and contacts in the power supply circuit to the control rod drive motors. When the brake circuit is de-energized, control power to contactors K48, K49, K50, and K51 is lost, which causes their associated contacts to open, disconnecting power (120V) to the control rod drive motors. This causes any control rods which were being driven in or out, at the moment the scram occurred, to fall into the core. Contactors K48+K51 are manufactured by ITE, Model A103C. They are listed in the safety related subtier component list as Subt-499, seismic type 1, and environmental I.D. 1 (Class I not required for safe shutdown). 4 ~ 22 -

' I'. i . y,c i ' <E Table 7.1-2 Scram Parameters 0;;!. mer.si ran me as "a SensedWertebM Type and Ihusser of lagut Ostetter tesettee beelt teg$t Lead Wales Trly tegel e4 le. Itemmet Ileadmultch(1) Castrol team I of 1 tend (l-43) tb. Nemmel seemd nutsches (3) eestrel Board B.40 t of 3 8 18 cenets/ set 2. asutres emmetrees. mesleer Chamaele I. Il PCEWllell I ef I -f ( m ely u w t i 3. ante of mostres tw clee - ble insche chamaels lit. IV, V pcar itell a of 3 e I decease s amendessmen l (see only at Start e )4 (wide reage). per see 4. an.4een fles - blgh merleer theensle III. IV. V. PERf tiell I of 3 100E pesar 1488 peuer VI. Vll. Vill .~ t of 3 l 3. Primary coolant moisture. high esmoeletmentterI(8) Pter penetretles 3 et 3 plus e -GST 47 4 I of I er esigelet dampelet I of I hip level a 8. asheet atoes temperature. bl p Theruecouples (II) Reacter Butidleg I of 3 100tT 1875 % (4 theressegles are sembleed S e for I serem chaemel) 'r i j 3. peleary caeleet pretoure - h Presserefressmettersh(3) PERW phastretten a of 3 MO pole 80 pst bek reted j (see sely at reser)1 programmed with lead l S. Primary ecolant preneure high PressereTrenamittersI(3) Ptat Pometretles I of 3 Me pole 71/25 shove seruel i preneure programmed l We hd i 9. act ruheet lies pressure - 1er Pressure sultches (3) Turtles Buildlag 2 of 3 Sie pelg 35 psig j (see sely at Pomerle T le. Saoerheet lies pressere. les Pressure outsches (3) Turtles But1dlag 2 of 3 2000 psig 1900 pelg (see sely at Pomer)t i 11. Phat electrical system poner. thiderseltage relays (9) 4eer suen a of 3 4eg eelts eget huses IA. 30. 8 K less me, la Is. 8 K (testof3phasesse (2 set of 3 huses) less et soltage for 35 seceede 3 12. Ime-leap treettei temp shutdeias logic teatrol Room 2 of 3 M 4 ] femord 1 10) (bothleaps) g i 13. asector t.lldlag eenporeture. Ther. ace.in (3) weter emeldeeg 3 of 3 es'.WF 3es*r i 1 hip tra l N Instotles le parenthesis refers to laterlech Sequence Sutlch er Reacter IInde $ witch posttless. h same trenametters are used for staaetuoter esp. s i.e.les, w.mie se e sendet$ oorek, e= steen genere er h., is e.ummi ad tr mie met ussN seeuelly cause a leap shutdans is sensed le tem eter steen generater leep, ktast erly setpolets are aere esmeerestive to allow for testriment.feescuracy. l we 1

7 ATTACHMENT 2 Fort S't. Vrain #1 Technical Specifications 4 Revision 13 - 6/29/76 Page 4.4-3 i smiff ection LC0 L.L-1 TABLE k.k-1 TNSTR'JMENT 0?CuTI:3 RE.JIREJC TS 70M ?LA.*:7 r.w.:c_xIVE SYE"'D-!. S.7a*4 MIND 1M MIKIM.t4 FI!D!ISSIBLE TRIP OPIRA3LE DEGREE OT BIPAES Wo. FUNerIONAL UNIT STITIPJ C.vAUNELS REITJ LANCY CONDITICH3 la. Manual (Control Roc =) 1 0 Nome 2 (f) 1 None Ib Hanual (hergency Board) 5 2. Startup Channel-Eigh 1 10 eps 2 1 Reacter Mode Sv. in "RUH" 3a.. Linear Channel-High, e ikOI Power 2 (f) 1 None-Channels 3, b, 5 Ta) 3b. Linear Channel-Eigh, e ikO5 Power 2 (f) 1 Kone. Chana=1s 6, 7, 8 Ta) J l Primary Coolant Moisture Rev Eigh Level Monitor 167'F Deypoint ICf,tl 1 (e) -None y,* Icop Monitor 127 F %. point 2/ Loop (f,tI 1/Icop (h) 8 g 5'. Reheat steam Temperature 1 1075'r (a) 2 (b) (f) 1 None - Eish (b) 6. Primary Coolant Pressure 1 50 psig below 2 (f) (k) 1 Less than 305 - Lov normal, load rated power programmed (a) 7 Primary Coolant Pressure 1 7 5% above 2 (f) (k) 1 sone d - Eigh normal rated. leed programmed (a) i 8. Ist Reheat Header 2,35 peis 2 (f.) 1 .Less than 3C5 Pressure - Lov rated so er 9., Main Steam Pressure 2,1500 Psig 2 (f) 1 Less than 305 - Low rated power

10. Plant Electrical, (d) 2 (e) (f) 1 None System-Loss 2

1 Reactor mode

11. Two Loop Trouble switch in f

" Fuel Leading"

12. Rich Reactor Building 5 325'F 2 (f) 1 None Temperatura (Pipe Cavity).

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m l' ' 2. m ime. 11 8 I Table S.4-1 l II sleest pasquusciss rea . -aaaensus. AIID 75873818 er SCIWIsTsist Cheeeet Deess8 ellen Famettes Freemenes 813 g 1. 88mment (Centrol Room) e. Test R e. Itemmelly trip eyetse i-1. Ilmamel (1-49) e. Test N s. llennelly trip each ebennet 3. start 1

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e. .o.,eri e., tee.e,. cot. e - 1. die.t.r. I b. Teet F b. Ieternet test signet te verify tripe, and eterne 8 f. e. Celtbeste a e. Internet test elseet she11 he eheeled and eellbrated to essere that its eetput le la eteerdance with the deelse requireusets. This shall be done efter eeuplet-I a ing the outernet teet etenet preeedere by ebeektes the 1 eutput todiestten etse terates the Seternet test signal setteh. I 4 e 4. Lineet pause Chenmot e. Cheek 3 e. Comperison of 6 seperate ehennet indlestore i I i II b. laternet teet signet te verify tripe, and eterne b. Test i 8 ,I e. Cottbrete B e. Chemmet adjested to egree with heet belemme seleetettee f d. Cottbrete E d. Interest Test olpelo to adject tripe end todiestiene i 5. WSee Roses power thenmal e. Etnek 3 e. Camperleen of three coperate tedlestese t b. Teet p b. Intesnet Test signets to verify tripe med eieres 1 e. Calibrate Il e. Ehemmet adjested to agree with heet betones seleetetten I'. 3, d. Smitheate a d. 1sternet Test signets to adjeet tripe end Sedicattene u 3= Cooperleen of tee coperate high lesel eheneet sterer 6. Frlmery costaat Isoletere e. thoek 3 e. m> i i (ett ehemsele) temperatore fedteettene j [ * $, w h MkHM b. Cheek 9 b. Camperteen of ein seperate low lesel ebennel sterer 2 QQ temperatore ledteettene eg r j w2"m b8 w b " " et 'a m m es UT

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.n -.d ?.. . s.. ~ - = = ATTACHMENT 3 PAGE 3 Fort St. Vrain il Technical Specifications Amendment 25 - 3/2/82 Page 5.4-5 i -l

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.c e ATTACHMENT 2 TO P-85040, FORT ST. VRAIN CONTROL R00 DRIVE AND ORIFICING ASSEMBLY REFURBISHMENT PROGRAM PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION PREPARED BY: Ken Ewell Proto-Power Corporation APPROVAL: 4 Frank 'Q Novachek ~ Technical / Administrative Services Manager e l l

FORT ST. VRAIN STATION CONTROL R0D DRIVE AND ORIFICING ASSEMBLY REFURBISHMENT PROGRAM PROGRAM DESCRIPTION Introduction This report describes the program currently being undertaken at Public Service Company of Colorado's Fort St. Vrain Generating Station to refurbish the reautor control rod drive and orificing assemblies (CR00A). The report includes both a description of the CR00A components to be inspected, tested, and refurbished or

replaced, as neces sary, as a part of the program and a description of the procedure to be used for disassembly and reassembly of the CRD0As.

Replacement parts to be used will either be manufactured to the original equipment specifications or be an upgraded design to resolve problems which have been experienced or anticipated. Where upgraded parts are to be used, the changes in design have been demonstrated to be suitable for the intended applications and doc ~umented by existing design change procedures. The overall purpose of the CRD0A refurbishment program is to ensure both that the CRD0As will perform their intended safety functions and that potential operability problems with the CRDOAs will not limit plant availability.. b

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SPECIAL EQUIPMENT FOR CRDOA REFURBISHMENT 1. Carousel with Rod Tubes, Clevis Holders, and Valve Support Stand - HSF(W) 2. Shield Wall with Lead Glass Window, Manipulator, and Clevis Wrenches - HSF(W) 3. Hydraulic Cable Cutter - HSF(W) 4. Clevis Cask and Cart with Track - HSF ~ 5. TV Cameras - HSF(W) 6. Decon Brush Ring - HSF(W) 7. Rotatable Shield - HSF(E) 8. Access Platforms - HSF(E) 9. Special Lighting and Power Supplies - HSF 10. Communications System - HSF 11. Special Ventilation: HEPA Unit and Ducting - HSF 12. Airlock and Special Access Door - HSF 13. CRDOA Support Stand for Baron Ball Removal - HSF(E) 14. Boron Ball Container - HSF(E) 15. Alignment Fixture for SA - HSF(W) 16. Baron Ball Removal and Fill Tools, including Air-Driven Vacuum Cleaner - HSF(E) 17. 10-Ton Gantry Crane with Rails 18. 1-Ton A-Frame Hoist 19. Transfer Shield with Bellows, Lifting Frame, and HEPA Unit 20. CRD0A Support Stands (upper and lower) for Disassembly - ESW i -

j CRDOA Components Involved in the Refurbishment Program The following CRDOA components will be inspected, tested, and refurbished or replaced, as necessary, as a part of the i refurbishment program: Component l Refurbishment Activities I i

1. Control Rod Drive l

Assembly l (200) Assembly l l

a. Shim Motor &

l Test and rebuild or replace, as necessary Brake Assembly l

b. Bearings l Clean or replace as necessary
c. Gears l

Clean, as necessa,ry

d. Limit Switches /

l Test and replace, as necessary Potentiometers l (Replace components previously l identified to be faulty.)

e. Control Rod Cables l Replace
f. Seals l

Inspect and replace, as necessary 1

2. Orifice Control l

Mechanism j i

a. Orifice Control l Test and rebuild or replace, as necessary Motor l
b. Bearings

] Clean or replace, as necessary

c. Potentiometer l

Test and replace, as necessary

d. Gears l

Clean, as necessary

e. Drive Shaft & Nut l Clean, as necessary
f. Drive Shaft l

Clean, as necessary Housing i I

3. Rod Retract Switches l

Replace (with cables) 1

4. Cable Seals l

Clean, as necessary I

5. Control Rods l

Verify serial numbers l

a. Clevis Bolts l

Replace with Inconel bolts I

6. Primary Seal Ring l

Inspect and replace, as necessary 1

7. Reserve Shutdown

'l System l l

a. Baron Balls l

Replace

b. Rupture Disk l

Replace, as necessary

c. DP Switch l Test and replace, as necessary

~ -

I

8. Helium Purge Check l

Test and replace, as necessary Valves l l l In addition, the following design modifications will be made to CRD0As.as a part of the refurbishment program: a. Installation of new purge seals on the orifice control mechanism mounting plate to improve control of helium purge flow into the upper housing of the CRDOA. b. Use of Inconel in -lieu of stainless steel for control rod cables, cable end fittings, and cable clevis bolts to eliminate the potential for stress ~ corrosion cracking in these components. c. Installation of RTDs in all CRDOAs to monitor ~ temperatures in the vicinity of the control rod drive assembly and orifice control mechanism. d. Installation, when required, of replacement seal material for seals internal to the 200 Assembly and the primary seal. Refurbishment Approach The approach which will be used to refurbish the CR00As was developed to meet the following program objectives: Ensure that safe shutdown capability is not affected a. during refubishment work. .b. Minimize personnel radiation exposure and Refueling Floor contamination levels. 1 h. Ensure proper quality control and documentation. d. Minimize the potential for problems. ~ i

3. Remove the control rods and deposit in carousel, as follows: a. Lower rods into carousel rod tubes. b. Engage rod clevis holders. c. Rotate clevis and remove clevis bolts. d. Cut swaged eye from each cable and deposit in cask. 4. Move the CRDOA from HSF (West) to ESW using 10 ton, gantry crane with Transfer Shield and position on ESW stands..(Upper stand raises CRD0A sufficiently to allow access to openings in the side of the upper housing for removal of' control rod drive assembly. Lower stand supports the orifice valve assembly during disassembly of the CRD0A.) 5. Disassemble the CRD0A. a. Disconnect electrical connectors and tubing through access openings. b. Remove CRD Assembly (200 Assembly) and place 4 in cart. Move to CR0 Refurbishment Area. c. Remove and dispose of control rod cables, d. Remove rod retract switches. e. Remove orifice control mechanism.

Inspect, test, clean and refurbish the mechanism, as necessary, f.

Remove the upper _ housing. Inspect and replace the primary seal, as necessary. g. Remove the shield container. h. Remove cable seals. Disassemble and clean for 4 reuse. 6. Refurbish the Control Rod Drive _ Assembly (200 Assembly). a. Disassemble and clean parts. 4 - 4 5 *

.. Attachment 2* CRD0A Refurbishmunt Program Hot Service Facility Special Shielding and Equipment . /* W W g,P e* d s immessf g e m fV ~* ,~ . (- = ',,;.I q <.* 3 1 .e,,, 4 , r q-i .a..,,. .,.4.. x.? s-g .s. ~

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