ML20034A104

Nonproprietary Licensing Rept for New Narrow Range Temp Measurement Sys,(Rtd Bypass Elimination)

ML20034A104
Person / Time
Site:	Callaway
Issue date:	04/12/1990
From:	UNION ELECTRIC CO.
To:
Shared Package
ML19302D976	List: ML20034A101 ML20034A103 ML20034A104 ML20034A105
References
NUDOCS 9004200104
Download: ML20034A104 (43)

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Section-Title ~

1.' O '-

LINTRODUCTION a

2.0

- BACKGROUND

3.' O OVERVIEW'OF=THE PROPOSED SYSTEM-~

13.1' Mechanica1LChanges

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3.'2 RTD Design _,

3. 3 ~ Electronic Modifications 1 7

3.4 System Accuracy.

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- 3.5.

ALARA Benefits

~3. 6 - RTD-: System' Time ' Response

4.0 DESCRIPTION

~OF MECHANICAL ~ MODIFICATIONS 3*

4.1 Hot Leg 4.2 Cold Leg 4:. 3 Crossover Leg 1

4.4 Inspection, Welding.and-Hydrostatic. Test Requirements-

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4.4.-l' Hot Leg and Cold LegiThermowells-4.4.2 Crossover Piping

. 2 4.4.3 _ Hydrostatic; Test' Requirements.

4.5 Analysis of-RCS' Penetrations' 4,6 Debris Control During Modification-

5.0 DESCRIPTION

OF ELECTRICAL / INSTRUMENTATION MODIFICATIONS 5.1 T-Hot Averaging

'I 5.1.1 Existing System 5.1.2 _ Proposed-System 5.2 T-Cold Monitoring 5.3 Weed:RTD.

5.4 ~.In-Situ Testing 7

5.5-Equipment Qualification

5. 6L Detection of a1 Failed RTD 6.0 ALARA-6.1 Description c;

6.2 Dose Savings 6.3 ALARA Methods.

6.4 Radioactive Waste-1 g

6.5 Radiological

Problems and Dosimetry i

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7.3 -Relocatio'nioflRTD

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- 7;4' Reactor Coolant - SystemE Flow:

-7.5

-SetpointLStudies

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l.0 INTRODUCTION x

The existing.RTD bypass piping.is scheduled to be removed during the 1990 refueling outage in September

-l at Callaway Plant.- A new Narrow Range Inline Thermowell Mounted' Reactor Coolant System (RCS)

Temperature = Measurement system willibe-installed during the same outagerin lieu of the~ bypass-system.

Combustion' Engineering (CE) has been selected by Union Electric to perform the' detailed engineering and-installation ofLthe-new system.

This report is submitted in support of continued operation of the Callaway Plant with the new RTD system installed.

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2.0 BACKGROUND

The current method of measuring the hot-and cold leg i

reactor coolant. temperatures'uses the RTD bypass system.

This system was designed to~ address -

i temperature streaming in the hot _ legs and to allow t

replacement'of the direct immersion RTDs without t

draindown of-the RCS. _-For increased accuracy in-i measuring the-hotileg temperatures,. mixing scoops are located inieach hot leg at three locations of a cross section, 120 degrees apart..Each scoop has five orifices which sample the-hot - leg: flow. :The flow.from the scoops is_ piped'toLa manifold where a direct r

immersion RTD measures the hot leg loop temperature upstream of the steam generator.. Thel cold leg temperature is measured-in a similar manner with piping to a separate bypass manifold,,except that no scoops are used as temperature streaming is not a problem due to.the mixing action of the RCS pump.

The resulting.

system consists of nearly 400 feet'of Reactor Coolant-

i Pressure-Boundary-(RCPB) piping, 64 associated' valves, 85 hangers which include 59 snubbers, 8 sets of flanges and 8 RTD manifolds.

Plant experience has demonstrated-9

-two major drawbacks to this design:

Lack of Reliability, - Plant shutdowns'have been Forced Outages required ~because of leakage (from valve packing or mechanical joints) or because oof flow reductions due to valve problems.-

High Radiation Dose -

The RTD Bypass Piping (B/P)

System is a significant-contributor to man-rem s

exposure because the numerous valves and. socket-welded pipes are crud _ traps.

Man-rem'is expended not only in maintaining and inspecting the RTD B/P System but in performing any work near the RTD'B/P System such as Steam i

Generator and Reactor Coolant Pump maintenance.

These problems a2.e not unique to Callaway but appear to-be common to all plants with a RTD B/P System.

The proposed narrow range, RTD System eliminates all the bypass piping, and its associated problems, while maintaining a fast response time, accurate hot leg temperature determination, and the capability to replace RTDs without draining down the RCS.

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a overview of-the new system is provided in Section 3.0.

Detailed descriptions are provided in Sections 4.0, 5.0, 6.0 and 7.0.

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3.0 OVERVIEW OF THE PROPOSED SYSTEM 3.1' Mechanical Changes:

All_the bypass. piping, associated valves, and RTD manifolds will be removed.

The three mixing' scoops in each hot leg will be retained.

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The top of the hot leg mixing scoops will be' modified-to allow welding,in a thermowell.

The1thermowell becomes part of the RCPB.

The RTD nipple and pigtail assembly screws into the thermowell (three RTDs for each hot leg).

The use of thermowells allows RTD replacement without draindown.

The nozzle on the cold log will be modified to allow welding in a thermowell.

The cold leg configuration is simpler because the Steam Generator and RCP provide adequate mixing of the fluid in that piping.

The crossover leg connection, through which the RTD B/P System fluid is returned to the main RCS piping, will no longer be required and, therefore,.

will be capped.

3.2 RTD Design:

Weed Instrument Co.,

Inc. dual element RTDs will be used in the new design.

Each RTD element will be shop tested inside a thermowell to ensure that the time response of both elements is within the required time.

Response time of the RTDs will be verified in the field using Loop Current Step Response (LCSR) methodology.

RTD accuracy will be-improved over the accuracy of the present RdF RTDs.

The spare RTD element will be wired to the 7300 Process Protection System cabinets (hereinafter referred to as 7300 cabinets)-so that switchover to the spare element can be done from the Control Room.

3.3 Electronic Modifications:

Each of the three T-Hot RTDs per loop will be wired up to an RTD amplifier (R/E-converter or NRA) card and the three signals then averaged to 1 !

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I produce one T-Hot signal which will replace the:

loop's T-Hot signallof-the existingLsystem. 'The added electronics will be identical to the existing 7300 electronic hardware now used.

Figure 5-1 shows'the concept and; outlines the added modules required.

3.4 System Accuracy:

ll CE has compared the temperature measurement accuracy of the new system with that of the old system, using Salem post-modification scoop temperature data'with CE ccoop test results.

In addition,-PSE&G compared T-Hot before the.

modification with T-Hot after the modification on Salem 1 and 2 using~100% power calorimetrics.

j Agreement was obtained_on.both units within 'he-c accuracy-of the data..-CE has' established the temperature n,easuremont bias as [-

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By retention of the mixing concept at the hot leg scoops, the sampling performed by.the existing system and, therefore, the process measurement cccuracies will be preserved.

ALconservative

.j temperature measurement bias of 0.27 F has_been j

included in the setpoint calculations discussed in j

Section 7.5 to reflect hot leg streaming effects.

Secondly, the electronic modifications represent j

an accuracy improvement, due to the: use of three

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hot leg RTDs and their separate RTD amplifier i

cards, that virtually offsets the effects-of:the hot leg streaming bias and the revised Delta-T q

gain discussed in Attachment 6 (reflects actual j

100% Delta-T span at Callaway).

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The same rack functional checks and calibration accuracy requirements will be maintained.

' Thirdly, the proposedLWeed RTD has an overall i

sensor accuracy, shown in Table 5-1, which is an improvement over the existing.RdF RTDs as-

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discussed in Sections 3.1.1 and 3.1.2 of.

3.5 ALARA Benefits:

A '2000 man-rem dose savings is ' projected over the remaining-life of the plantias-a result of this modification assuming a 40-year Operating License, i

This ALARA:and' cost benefit analysisitakes_into

= account thefreduced~ radiation levels, reduced

-outage time, increased = accessibility in the loop

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compartments, installation / demolition doses, maintenance requirements and the plant's reliability over the-life of the plant.

3.6 RTD System Time Response:

The RTD system response-time is the time lag from-when the hot leg temperature reaches trip conditions at the scoop until the control rods start to drop into the core.

The Technical Specification (TS) Required Response' Time includes-only that portion which can be: tested.

At Callaway--this is six seconds.

The testable response time of the new system design, listed in

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Table ' 7-1, will be lese.than -six seconds.2 The-total system response time is' reduced from that'of the current system through the reduction of loop travel and thermal lag time from.the scoop inlet ports to the RTD.

Since the response time of-the new system is less than the TS requirement, no TS changes to, response time are necessary.

The. time response used in the FSAR safety analyses and in WCAP-10961-P, "Steamline Break Mass / Energy Releases for Equipment Environmental Qualification Outside Containment," is eight seconds and, therefore, will also remain unchanged:(refer to FSAR. Table-15.0-4).

Response _ times listed in Technical Specification Tables 3.3-2 and 3.3-5 will continue to be verified every refueling outage.

4.0 DESCRIPTION

OF MECHANICAL MODIFICATIONS' 4.1 Hot Lea:

1 The hot leg installation has three nozzles, 120 apart, around its circumference.

The nozzles 1

extend into the pipe to form scoops to sample the flow.

The scoops will be' retained in the new,

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design and will collect a_ flow sample;in a manner -

equivalent to the present configuration.

A thermowell will-be mounted inside each scoop.

The scoops will be modified so that.the flow goes past the thermowell (Figure 4-1).

Since the existing sample scoops are being rettined, the method of sampling tthe. stratified flow in the hot legs will' remain unchanged.

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The portion of overall-response time attributable to the flow through the scoop is 0.25 seconds or less.

This value includes fluid transit and heat capacity effects and was conservatively estimated using:

t 4.2 Cold Leg:

The cold 11eg has a single' nozzle without a flow i

sampling scoop.

The nozzle is a 2" IPS socket-weld nozzle.

The cold leg thermowell will be installed directly into the nozzle (Figure 4-2).

As is the case with the present_ system,-no flow sampling will be necessary because'the Reactor Coolant Pumps will provide mixing of_the flow after it exits the-steam generator.

4.3 Crossover Leg:

The return for the bypass loops is a 3" nozzle in 4

the crossover leg.

This connection.will no longer be used._

A 3" schedule 160 buttweld cap will be installed on this connection (Figure 4-3).

4.4 Inspection, Welding and Hydrostatic Test Reauirements.:

4.4.1

' Hot Leg and' Cold Leg Thermowells The following-requirements are applicable to the 12 hot leg RTD scoops and 4 cold leg connections:

1.

Liquid penetrant inspect all accessible field machined surfaces in accordance with the 1980 Edition through Winter 1981-Addendum of Section-XI of the ASME B&PV Code.

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Welding to be in accordance with-the 19801 Edition through Winter 1981 Addendum of-Section XI of the-ASME B&PV' Code.

(Root' Pass-GTAA, Fill-GTAA or GTAW)-:

3.

Liquid penetrant l inspect'the root

-weld pass in accordance with the

-1980 Edition through Winter 1981 Addendum of Section'XI of"the ASME B&PV Code.

4.

Liquid penetrant' inspect final' weld

. pass in accordance with the 1980 i

Edition-through' Winter 1981 Addendum of~Section XI of the ASME B&PV Code.

5.

-Weld material to.be supplied in j

accordance with ASME<Section-II l

with additional requirements of ASME Section III'NB-2400 (1974 Edition through Summer 1975 Addenda).

4.4.2 Crossover Piping The following inspection and welding requirements are, applicable to capping-j of the 3" cros'sover. piping 1at four-

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locations in addition toLthe five i

requirements noted in Section 4.4.1:

I 1.

Radiographically inspect the completed weld.in accordance with i

the 1980 Edition.through Winter l

1981 Addendum of'Section XI of the ASME B&PV Code.

2.

An open butt weld configuration will be used with argon as purge j

gas.

4.4.3 Hydrostatic Test Requirements Hydrostatic testing of all nozzles will' be done during inservice testing in accordance with the 1980 Edition through-Winter 1981 Addendum of the ASME Section XI IWB 5000.

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4.5 Analysis of RCS' Penetrations:

The thermowells are pressure boundary parts which completely enclose the RTD.-

They will be-machined from a solid bar of SB-166, a nickel-chromium-iron-alloy and will be shop.hydrotested to 1.25 times j

the RCS design pressure. -The external design pressure and design-temperature-will be the reactor coolant system design pressure and f

temperature.

The RTD, therefore, will not be part of the pressure boundary.

For both the hot. leg and cold leg, the nozzle, thermowell, and the entire thermowell/ nozzle l

assembly will each'be analyzed to.the ASME'B&PV.

1 Code,Section III,-Class-1.

The. analysis of the entire assembly will consider the weight of the l

RTD, the-connectors as applicable, and an assumed length.of cabling.

The effect'of seismic and flow-induced loads will also be considered.

Seismic response spectra enveloping Callaway will l

be used.

Flow-induced vibration will also be evaluated.

This' stress analysis will be completed by July 1, 1990; however, based on CE's past experience, code allowables will be satisfied.

1 The crossover. leg connection will be analyzed-to the same requirements as the hot and cold leg connections.

Since the connection will be capped-and have no piping loads, stress levels will be-lower than what exists in the current system.

4.6 Debris Control Durina Modification:

Control of metal chips and fragments.will be as follows:

Hot Leg Scoop Modifications:

(12 locations)

For those locations which tre done with the system full of water, a freeze plug will be installed l

prior to any cut-off operation.

A mechanical plug will be installed after the freeze plug is removed and prior to any machining which would develop metal chips or fragments.

Any chips or fragments-will be removed by vacuum prior to removal of the mechanical plug.

The holes in'the scoops will be made using the Electrical Discharge Machining (EDM) process.

EDM is a process that utilizes electrical discharges, or sparks, to machine metal.

The surface being machined is bombarded with high intensity electrical energy pulses that gradually l

melt away the stock until the desired configuration is obtained.

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A high' energy spark, through: vaporization,-

melting, and an explosive effect,' dislodges a:

minute particlefcf metalt from the workpiece, t

leaving a small crater.

The. dislodged particle is-then solidified and washed away by the dielectric fluid.

Some or all of the four top 1 scoops may have the pipe cut-offLoperation performed-at part-loop drain making a freeze plug unnecessary.

Cold Leg Nozzle:

(4 locations) o The cold leg nozzle w111.1x3: cut-off after-draindown of the.RCS, A. barrier will be' installed.

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prior to cut-offsto catch chips.or fragments which will then be removed by vacuum.. There will also; i

be a barrier in place during weld preparation-machining of the nozzle.

Crossover Leg Nozzle:

.(4. locations)

The primary system will be~ drained prior to any cutting operation on the crossover leg nozzle.

The cutting operation will1be performed using an abrasive cut-off wheel which does not. develop metal chips.

A mechanical plug will be installed prior to machining of the' weld preparation and the nozzle vacuum cleaned prior to removal.

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5.0 DESCRIPTION

OF ELECTRICAL / INSTRUMENTATION MODIFICATIONS

- 5. ll T-Hot Averaginar I

5.1.1 Existing System-The fluid from the three scoops for each loop is mixed together before being-directed to the~T-Hot RTD bypass manifold.

At the manifold, a single RTD is_used to measure _the temperature._ The

- i resistance'across this RTD is connected to-a resistance' bridge.

The resultant differential voltage is amplified to provide an amplified voltage output (R/E) before being; combined with the T-Cold signal to generate the loop's-Delta-T and T-AVG. signals used by the 7300 cabinets.

Refer to' Figure 5-1, 5.1.2 Proposed System The proposed system will locate a dual element RTD in each of:the'three scoops.

Averaging'of-the RTDs at the three locations ~will be done electronically.

Refer to Figure 5-1.

The~ resistance across one RTD element at'each location will be connected'to its respective-RTD amplifier.

The: amplified signal. from the three RTD amplifiers will.be averaged together to generate a single T-Hot signal for that loop,'T-HAVG, which along with the T-Cold signal is then used to generate the loop's Delta-T and T-AVG signal.

The second RTD element at each location is considered an installed spare.

They-will~be wired up-to the' Master Test (NMT) cards in the 7300 cabinets, but not normally connected to'the RTD amplifier cards.

On failure of the first element, the second element is available.

(Refer to Section 5.6.)'

The existing RTD amplifier cards (R/E converter or NRA) are designed to accept 3 wires.

The 4th RTD lead wires.are connected to the NMT cards in the 7300 cabinets...

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5.2 T-Cold Monitoring:

The impact on the T-Cold portion of the system is limited to:

(1)

Relocation of the'RTD from'the bypass manifold-into a thermowell directly-in the RCS cold leg piping.

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(2) -Use of one dual element RTD instead of two single element RTDs.

.As with'the hot. leg i

RTDs, both elements will be wired toLthe 7300 cabinets but only one element will input into the RTD amplifier cards.

As before, the RCS. fluid is-mixed-by the Reactor Coolant Pump before reaching the cold leg.. No sampling scoop =is required.

The intended location of the cold leg RTD/thermowell11s the nozzle used as a tap-off point for the existing. cold leg RTD bypass line.

5.3 Weed RTD:

L Dual element RTDs have been supplied and;are-in

.use at other operating plants-including Waterford Steam Electric StationtUnit-3 and1SalemEUnits 1 and 2.

.The RTDs are provided'with' Resistance vs.

Temperature (R vs. T) calibration curves which are accurate to a-specification of 1 0.3'F.

The RTD drift is specified to be within i 1 F over-a five (5) year period.

See Table 5-1 which.is based on the drite term being linear with respect to time, The Weed RTD initial shop calibration is performed D

by immersion in ice and oil baths.whose f-temperature is monitored by a standard RTD calibrated'to NBS standards.

The RTD/thermowell response time is measured by plunge method by causing a' step change from ambient' room-temperature to elevated temperature..All RTDs must meet a specified. response. time requirement and will, therefore, be interchangeable.

The dual element design provides an installed i.

spare wired up to the 7300 cabinets for use when prir.ory element failure'is detected.

For this reason, both elements of each RTD will be' tested by Loop Current Step Response (LCSR) for.in-situ response times on a refueling outage interval after installation..

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The contact between the RTD and thermowell is a critical item in maintaining the response time.-

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The above-1 provides high assurance of consistent response time.

l 5.4 In-situ Testing:

.The Weed RTD is capable of being tested by the in-situ'LCSR method.

A continuous current of 20-40mA will not damage the RTD.

5.5 Equipment Qualification:

The Weed model N9004 is qualified to IEEE 323-1974j IEEE 344-1975 and-NUREG-0588/10CFR50.49 to levels'.which envelope'Callaway Plant requirements,-as specified in FSAR Sections 3.7(N), 3.9(N), and 3.11(B),

Based-on an.RTD service temperatureoof 135"F'at the epoxy-sealed RTD-lead wire transition, to'be established by 4

shop testing with consideration'given to process fluid heat rise effects,7a qualified life of 40 years has been established, The 4-wire dual element Weed-RTD to be used in this modification is qualified by similarity _to the combined features.of two of eight.RTD test specimens in Southwest Research Institute Report-

1. 06-8680-003.

One of those two, a fast response i

dual element 3-wire RTD/thermowell assembly, also featured a sealed NEMA-4 head.

The other~ specimen' taken credit for is a direct immersion RTD assembly with Swagelok fittings and 316-SST flexible tubing, seal welded or brazed to the fittings.

Callaway will use a 4-wire, fast response dual element RTD/thermowell assembly-(no head) with the required fittings and sealed flexible tubing necessary to mate up with the quick disconnect assemblies discussed below.

The 4-wire leads are made of the same insulation and jacket as that tested.

The test RTDs were calibrated after the LOCA simulation and demonstrated only a i O.1 F change from the baseline calibration data.

See Figure C-2 for a schematic of the RTD assembly.

Containment Penetration Module Assemblies:

The Bunker Ramo 85#16 AWG and 20#16 AWG penetration modules with shielded twisted quad pigtail assemblies are-qualified for Callaway applications, including-IEEE 317-1976 and IEEE 323-1974, as discussed in ULNRC-1992 dated 4-27-89 and as documented in Wyle Labs-Test Report-

1. 17040-1 performed on-two spare penetration module / pigtail assemblies obtained from Callaway.

RTD Quick Disconnect Assemblies:

The Patel/EGS Quick Disconnects are supplied by Weed as part of the RTD assembly.

One connector half will be connected-approximately 5 feet from-the RTD and be made an integral part of.the RTD assembly..The other connector half will be.

provided with a 25 foot-pigtail having a' qualified Patel/EGS conduit seal'with #20 AWG wires for splicing to field cable.

The assembly also i

provides an environmental seal.-for protecting the RTD lead wires from' harsh environments during accident conditions.

The Patel/EGS' connector assembly is qualified to IEEE 323-1974 and IEEE 344-1975.

The qualification levels' exceed Callaway. Plant requirements.

Two of the connector. assembly test specimens included Weed #20 AWG leads and used the Weed potting precedure.

1 Sealed Flexible Tubing:

A sealed flexible SST tubing will be installed between the RTD Quick Disconnect Assembly and the existing Junction Box.

The 3/4" SST flexible tubing is qualified to IEEE-344-1975.

The qualification levels exceed Callaway Plant requirements.

Field Cables:

The new field cable to be used for-this modification is Rockbestos 32/C, 8 shielded twisted quads, #16 AWG tinned copper, insulated and jacketed with flame retardant cross-linked.

polyethylene (FR-XLPE).

Each quad is. twisted with

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a #18AWGLdrain wire in contact with an aluminum / mylar shield.

This cableHis qualified to IEEE 323-1974 and IEEE 383-1974 by Rockbestos Test Reports QR-5805 and QR-7804 and satisfies Callaway Plant requirements.

Junction Boxes and Splices Current plans - call for existing junction boxes e

to be used to house the splices between the extension leads from the quick disconnect assemblies and the field cable.

These junction boxes are NEMA 3R, Hoffmann terminal. boxes with low point drain holes.

If needed, additional junction boxes of this type will be added.:- The conductors to be spliced will.be' joined by either ring tongue terminals bolted together-(Wyle-Labs Test Report #58722-2) or'by Burndy type YSV o(or approved equal) butt splices (Wyle Labs Test-Report #58442-1) and then' sealed with Raychem WCSF-N heat shrink tubing.

This splicing methodology is also.used'at the containment penetration and is qualified to IEEE 323-1974 and i

IEEE 383-1974 for Callaway Plant requirements.

Reactor Protection System (RPS). Hardware:

The added test cards (Temperature Channel Test, NTC and Master Test, NMT), RTD amplifier cards (R/E converter or NRA), summing amplifier cards-(NSA), and isolator cards (NLP) are identical to the existing-7300 electronic' components used-at Callaway.

The electronic module used to derive the loop's average T-Hot signal (T-HAVG) from the individual T-Hot inputs is identical to the module now in use to derive the loop's average temperature _-(T-AVG)

L from the T-Hot and T-Cold inputs.

The added electronics will be installed in spare card locations in the existing 7300 cabinets.

Existing divisional separation will be maintained.

All additional mounting hardware will be identical to existing mounting hardware.

All new electronics and mounting hardware will be procured through the same source used to supply the-present equipment.

Cabinet wiring will meet Class lE E

requirements.

The 7300 Process Protection: System cabinets were qualified per WCAP-8687, EQTR E13A through E13D, to IEEE 344-1975.

The 3-bay test cabinet was loaded with cardframes, power supplies, and dummy weights to simulate possible loading extremes that

i l

t could exist.. The dummy weights included an=

allowance to simulate internal. cable weights.

Five OBE's and'four SSE's were simulated with fully loaded cabinets to_g-levels more than 3 times. higher than anticipated at Callaway.

The' additional mass due tx) new cards and-cables i

associated with this modification has been determined to be enveloped.by thisLtest program.

The above ensures that the added electronics will-be compatible with existing _ electronics.. It also minimizes the impact on present-training and procedures.

In addition, all the equipment has been fully.qualffied and has a demonstrated high-reliability.

5.6 Detection of a Fajjgd RTD:

A failed RTD would-pa~ picked'up by the loop.

Delta-T vs. auctioneered (high) Delta-T: deviation-alarm _ currently set at i 7.41% rated' thermal' power-and/or the loop T-AVG vs.;auctioneered1(high)

T-AVG deviation alarm currentlyLset at 1 3'F.

Also,.each channel is checked once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> as required by the Technical. Specifications.

On failure of an RTD, the: channel would be tripped and the Technical Specification Action Statement would go in effect.

Since'the Delta-T-protection functions require 2fout of 4 logic, the failed channel would have no impact cn1 the safe operation or shutdown of the plant.

As discussed in Sections 5.1.2 (HotELeg) and 5.2 (Cold Leg), the second element'of each RTD is:an

" installed spare" which._is wired all the way to the Master Test.(NMT) cards-in the~7300 cabinets.

This facilitates changing to the spare element as well as minimizing the time'that one channel would have to be tripped.

i

i 1

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TABLE 5-1 WEED RTD ACCURACY *_

ACCURACY (INCLUDES HYSTERESIS AND REPEATABILITY) 1 0.3*F l

DRIFT (@ 24 MONTHS) 1 0.4*F I

TOTAL UNCERTAINTY

+-0.7*F

• The more conservative RdF values were retained-in the new Callaway setpoint calculations in Attachment 6.

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CALLAWAY PLANT HOT LEG TEMPERATURES

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.7 6.0 AL&Rh 6.1

==

Description:==

The project will involve the removal of all of the RTD_ bypass manifold piping which consist of about 400 feet of piping and 64 valves.

Following the removal of the piping, existing penetrations into the RCS piping will be modified to allow the installation of the new thermowells.

-The major steps involved in installation of the hot leg thermowells are:

1)

Cut the' pipe stubs remaining.

2)

Prepare the end of the nozzle for acceptance of the thermowell.

3)

Using the EDM tool, bore out the new flow holes.

4)

Install thermowells and weld.

\\

The major steps involved in installation of the cold leg thermowells are:

1)

Cut the pipe-stubs remaining.'

2)

Prepare the end of the nozzle.

3)

Bore out the cold leg nozzle to accommodate-the new thermowell.

4)

Install thermowells and weld.

The major steps involved at-the cross over leg are:

1)

Cut the pipe stubs remaining.

2)

Prepare the end of the nozzle.

3)

Weld on the new caps.

The isolation valves and the RTD manifolds are the major sources of radiation in the existing system.

It is expected'that the removal of these components will be the most exposure intensive portion of the demolition phase. s

i a

The radiation exposure rates at the RCS penetration work areas will be reduced by the removal of the isolation valves and RTD manifolds as well as the installation of large quantities of i

temporary shielding on the RCS piping.

6.2 Dose Savinus:

The arrangement of the Callaway RTD manifold piping is such that the high radiation fields generated by them increase the collective exposure received during steam generator and reactor-coolant pump inspections or maintenance.

Although temporary shielding is used to reduce these radiation-levels during long outages, it is not used for forced outages as it requires at least i

1.5 days to erect.

Even with the use of temporary shielding, approximately half of the dose received i

during steam generator and reactor coolant pump maintenance is attributed to the RTD bypass piping which is in the same area.

The removal of the RTD bypass manifolds is expected to reduce the collective exposure by about 2000 man-rem over the remaining life of the plant assuming a 40 year operating license.

In addition, forced outtges will be avoided due to the avoidance of leaks and equipment failures.

6.3 ALARA Methods:

The project is being reviewed and planned in accordance with ALARA procedures.

Use of temporary shielding is planned for the modification of the RCS piping penetrations.

The use of respirators will be minimised by local decontamination and by the use of a containment system during the se.tchining process.

6.4 Radioactive Waste The waste generated by this project will consist largely of the removed piping and semi-encapsulated insulation.

Itgsexpectedthat a vaste volume of less than 800 ft (including piping, insulation, valves, and supports) will result.

The disposal method will be the usual low L

level waste burial process.

6.5 Radiolo_g_ical Problems _and Dosimetry:

Although some components of the RTD bypass manifold piping system do present rather high exposure rates, these can be managed by the use of the ALARA planning process.

Containment systems.

o I

i will contain almost all of the loose surface contamination and/or airborne.radioactdvity that might be produced during the machindnt process.

7.0 SYSTEM FUNCTIONAL IMPACTS The narrow range RTD temperature outputs.are used for a number of purposes including reactor trips and Engineered Safety Features (ESF) actuation, ms well as electrically isolated control systems, alarns, computer inputs, and indicators.

7.1 System Accuracy:

The effect on accuracy of the proposed-system.is insignificant because of the followings o

Hot leg scoop mixing has been retained as discussed in Section 4.1.

o The replacement RTD is specified to have an improved accuracy / drift over.the existing RTDs.

The accuracy of the new RTD is'.

discussed in Section 5.3.-

Since the new-RTDs will not be in contact with the primary fluid and will be provided with a quick disconnect at approximately 5 feet beyond the union along the pigtail, they can be-readily removed.

Little, if any, decontamination would be required to allow transport to a testing facility to check calibration of the RTDs.

L o

Each hot leg RTD will be wired to an RTD l

amplifier card (R/E converter or NRA) which is then wired with the other hot. leg RTDs in that RCSLloop to a summing amplifier card (NSA) which averages the three signals to obtain the loop's T-Hot.

1By having three parallel path T-Hot RTDs, R/E converters, and-interconnecting wiring,- the sensor errors-are noticeably reduced while'the rack errors are only slightly-reduced due to the added NSA card, as discussed in Attachment 6.

p 1

i r-o The impact of the T-Hot electronics (Figure 5-1) has been evaluated as discussed in Section 3.4.

The existing Technical Specification channel functional checks, response time tests, and calibration accuracy requirements will be maintained.

The impact of rack drift has been considered in the evaluation.

o There is no change to the cold leg's electronics; and therefore, no impact to the accuracy other than the benefit obtained from a more accurate RTD.

o These factors virtually offset all of the effects of the hot leg streaming temperature measurement bias and the revised Delta-T gain.

7.2 Epsponse Time Impact:

This modification will not impact the Technical Specification (TS) instrumentation response times.

This is because the total testable response time of the proposed system is shorter than the time specified in the TS as shown in Table 7-1.

With the proposed system, the response time of the RTDs will be determined with the RTDs in the thermowells (using LCSR methodology); therefore, the thermal lag associated with the thermowell will be included in the RTD tested response time.

The response time of the-proposed RTD/thermowell is 0.75 seconds slower than the existing direct immersion RTD's response time.

However, the fluid travel time from the inlet port of the scoop to the RTD is reduced from 2.0 seconds to 0.25 seconds with the new system.

OTDT trip time response is modelled in two parts.

The first part is a first order lag- (i.e., thermal lags and RTD response time) and the second part is a pure delay (i.e.,

electronics delay).

Depending on the transient and the OTDT equation, a first order lag can result in later rod motion than a' pure delay of the same magnitude.

Since the existing FSAR accident analyses and the Steamline Break topical (WCAP-10961-P) were performed with 6 seconds of first order lags and 2 seconds of pure delays, as shown on Table 7-1, no previous analyses are impacted since the total response time of the proposed system is faster than the current system.

c The times listed in Table 7-1 bound best estimate response times.

The allocated time for the RTDs includes a 10% error allowance for LCSR testing, 7.3 Relocatipp of RTD Instruments:

The function of the RTDs in the bypass piping i

manifolds is to measure the RCS hot leg and and cold leg temperatures.

Accordingly, physical relocation of the RTDs into thermowells mounted directly in the RCS piping is consistent with the function of the RTDs.

At the proposed locations, the RTD thermowells will be directly in.the RCS flow path and not have to rely on a subsystem.

7.4 Reactor Coolant System _ Flows f

Elimination of the RTD Bypass Piping System will i

have a very slight increase of approximately 0.1%

i in the flow through the Reactor Vessel and Steam Generator.

Although this flow increase theoretically results in better heat removal-at 1

I the core and better heat transfer at the Steam Generator, the change is too small to be significant.

7.5 Setpoint Studies t

The effects on setpoint terms Z, S,

and Allowable i

Value for the OPDT reactor trip and the Trip Time Delay (TTD) Delta T interlocks for steam generator low-low level reactor and ESFAS trip functions were assessed.using previously approved setpoint methodologies, revised in Attachment-6 to account-for the new hot leg RTDs and added electronica yet maintaining the basic approach (i.e. SQSS g

combination of independent uncertainties).

Only--

minor changes to the Technical Specifications are needed to account for a worst case bounding value for the hot leg streaming temperature measurement bias, discussed in Section 3.4, and to. reflect an updated Delta-T gain which converts temperature values in degrees F to the corresponding value in percent Delta-T' span.

This gain is based on recent plant data.

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TABLE 7-1 SAFETY **

EXISTING PROPOSED ANALYSIS I.

FIRST ORDER. LAGS a.

Direct Immersion 4.0 sec.*

N/A Bypass Manifold RTD E

Response Time

{

i b.

Combined RTD/Thermowell N/A

.4.75 sec.*

l Response Time

• j i

c.

RTD Bypass Line Fluid' 2.0 sec.

N/A Transport Delay and

.[

Piping Thermal Lag d.

Scoop Transport Delay Included 0.25 sec.

and Thermal Lag in c i

SUBTOTAL FIRST ORDER LAGS 6.0 sec.

5.0 sec.

6.0 sec.

II.

PURE TIME DELAYS a.

Electronics 1.5 sec.

1.0 sec.

b.

SSPS 0.010 sec.

0.010 sec, t

c.

Reactor Trip Breakers 0.150 sec.

0.150 sec.

SUBTOTAL PURE DELAYS 1,66 sec.

1.16 sec.

2'.0 sec.

r TOTAL TESTABLE TIME DELAYS ***

5.66 sec.

5.91 sec.

TOTAL TIME DELAYS 7.66 sec.

6.16 sec.

8.0 sec.

• Includes 10'f test allowance for LCSR testing.

Existing RTD response time makes use of~ time margin available in the OTDT analyses.

• • See FSAR= Table 15.0-4.

j

• • • Tech. Spec. limit is 6.0 sec. (excludes transport delays and thermal lags).

. 0

ULNRC-219 6

[

ATTACHMENT 2 SAFETY EVALUATION FOR RTD BYPASS ELIMINATION y

P i

SAFETY EVALUATION i

This amendment application includes revisions to Technical Specification Tables 2.2-1, 3.3-4, and 4.3-1 to accommodate the proposed replacement of the current RTD bypass system with an v

RID /thermowell-system incorporated directly into the hot and cold legs of the reactor coolant system.

This modification is desirable in order to increase plant availability and reliability due to the removal of several valves that have been the source of leakage inside containment and to reduce man-rem exposures in keeping with the objectivcs of our ALARA program.

As described in Final Safety Analysis Report (FSAR) Section 5.1, Reactor Coolant System (RCS) hot'and cold leg temperatures are measured by narrow range, direct immersion RTDs located in bypass manifolds.

Through the use of a bypass manifold around each steam generator, hot leg temperatures are obtained by mixing the flow from three scoop connections which extend into the flow stream at locations 120' apart circumferential1y.

Flow for the cold leg manifold is obtained downstream of the reactor coolant i

pump.

Both hot leg and cold leg bypass flows enter a common return line to the cross over leg (see FSAR Figure 5.1-1, Sheet 1 for the existing configuration),

i As discussed in FSAR Section 7.2, the existing RTD teraperature outputs are used for a number of purposes.

They are used by the Reactor Protection System for the Overtemperature Delta-T (OTDT) and Overpower Delta-T (OPDT) trip functions.

The OTDT reactor trip function is a primary trip credited in the accident analyses (FSAR Chapter 15) and WCAP-10961-P (Steamline Break Mass / Energy Releases for Equipment Environmental Qualification Outside Containment).

The OPDT reactor trip function provides backup protection against excessive power (fuel rod integrity protection).

No credit is explicitly taken for OPDT trips in the Callaway FSAR Chapter 15 accident analyses.

The steam generator low-low level reactor trip and AFW initiation circuitry has Delta-T interlocks that provide Trip Time Delays (TTD) at low power conditions.

Rod control is based upon T-AVG signals isolated from protection-system channels.

The T-AVG signals are also provided to the nressurizer level control system, the steam dump control system ('.n the T-AVG mode), control rod insertion limits, rod stops and turbine runbacks, and certain interlocks.

The functions that utilize temperature input from the existing narrow range RTDs will not be affected by their proposed removal and replacement because the signals derived from the proposed replacements will be equivalent to those provided by the existing RTDs.

The proposed change involves removal of the existing bypass lines and replacement of the existing RTDs with thermowell RTDs.

Three dual element RTDs will be used for each hot leg.

These will be located in the-existing scoops.

One dual element thermowell RTD will be located in the existing cold leg penetration.

The r

P nozzles in the crossover legs for the return lines will no longer be needed, and they will be capped.

The proposed replacement of the existing RTD and bypass line elimination does not involve an unreviewed safety question as discussed hereinafter.

t The removal and replacement of the existing RTDs will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR.

The consequences of an accident or malfunction of equipment important to safety previously evaluated are considered first.

There are six non-LOCA analyses of interests a)

FSAR Section 15.2.3, Turbine Trip (specifically the case with pressurizer sprays and PORVs available with maximum reactivity feedback) b)

FSAR Section 15.4.2, Uncontrolled Rod Cluster Control Assembly (RCCA) Bank Withdrawal at Power (specifically the case with a 1 pcm/sec reactivity insertion rate) c)

FSAR Section 15.4.3, RCCA Misoperation (specifically the case for a single RCCA withdrawal at power) d)

FSAR Section 15.4.6, CVCS Malfunction that Results in a Decrease in the Boron Concentration of the Reactor Coolant (specifically the case at full power with manual rod control) e)

FSAR Section 15.6.1, Inadvertent Opening of a Pressurizer Safety or Relief Valve (safety valve opening is limiting) f)

WCAP-10961-P, Steamline Break Mass / Energy Releases for Equipment Environmental Qualification Outside Containment These events are of interest because the OTDT trip is the primary trip assumed in the analyses, with a total response time of 8 seconds (6 seconds of first order lags and 2 seconds of-pure delay).

No OPDT trips are assumed in the analyses.

The OTDT, OPDT, and steam generator low-low level trips will continue to function in a manner consistent with the existing analyses assumptions for these events.

The total response time of the proposed system is less than that for the current system.

Further, the elimination of the RTD bypass system will not affect the LOCA analyses input and the results of these analyses will be unaffected.

Therefore, the plant design changes due to the RTD bypass elimination are acceptable from a LOCA analysis standpoint without requiring any reanalysis.

l l r-

L Hence, there will be no increase-in the consequences of an accident or malfunction of equipment important to safety previously evaluated.

.There will be no increase in the probability of occurrence of an accident or malfunction of equipment important to safety previously evaluated in the FSAR.

The events of interest are those initiated by a failure of those systems that use temperature inputs from the narrow range RTDs (T-AVG) or could be initiated by a mechanical failure of components affected by the proposed change.

There are four such events.

These are a)

FSAR Section 15.4.2, Uncontrolled RCCA Bank Withdrawal at Power b)

FSAR Section 15.1.3, Excessive Increase in Secondary Steam Flow c)

FSAR Section 15.1.4, Inadvertent Opening of a Steam Generator Relief or Safety Valve d)

FSAR Section.15.6.5, Small Break Loss of Coolant Accident (SBLOCA)

The Uncontrolled RCCA Bank Withdrawal event is an ANS Condition II (moderate frequency) event potentially initiated by a failure of the rod control system.

The Excessive Steam Flow and Inadvertent Steam Generator Depressurization events are also Condition II events.

They are potentially initiated by a failure of the steam dump control system.

The input to the rod control system and steam dump control system from the replacement RTDs will be equivalent to that currently provided by the existing RTDs.

The proposed modification will be done in a manner consistent with the plant design bases.

As such there will'be no degradation in the performance of or increase in the number of challenges to equipment assumed to function during an accident situation.

Furthermore, there will be no increase in the probability of failure or degradation in the performance of the systems designed to reduce the number of challenges to equipment assumed to function during an accident situation.

Hence, the i

first threc events will remain Condition II events.

The SBLOCA is an ANS Condition III (infrequsnt) event.

It could be initiated by the highly unlikely ejection of a thermowell or the failure of one of the caps that will cover one of the existing cross over leg penetrations.

All changes will preserve the qualification of the Reactor Coolant System pressure boundary.

The scoops, cross over leg buttweld caps, and thermowells will be analyzed to the ASME Boiler and Pressure Vessel Code,Section III, Class 1 and installed in accordance with Section XI of this code.

This stress analysis will be completed by July 1, 1990.

Based on CE's past experience, Code allowables will be satisfied.

The SBLOCA will remain a Condition III event.

Hence, there will be no increase in the probability-i E

l

of occurrence of an accident or malfunction of equipment important to safety previously evaluated in the FSAR.

Additionally, approximately 400 feet of small diameter piping and the associated valves will be removed from the primary system pressure boundary, thereby eliminating the possibility of a SBLOCA caused by a failure in this section of piping.

An increase in the RTD/thermowell response time will not increase the probability of occurrence of a previously evaluated accident or malfunction of equipment important to safety because the numerical value of response time is not the initiator of such an event.

The total response time of the functions that use signals from the narrow range RTDs will be reduced after the proposed modification.

The probability and-consequences of flooding and-jet impingement have been reviewed.

The thermowells and caps wi11 be in the same or immediate locations of the existing RTD bypass loop connections.

Therefore, the consequences of a postulated flooding-of the proposed RTDs or the impingement of a jet on the proposed RTDs are bounded by the results of exinting analyses.

There is no increase in the probability of flooding or jet impingement as the number of components and welded joints will be reduced considerably.

The consequences of a missile due to the postulated ejection of a thermowell has been reviewed.

The cold leg thermowells are considered first.

If ejected, these thermowells will impact either the RCP supports or the undersido of the operating deck floor with no appreciable damage.

A missile created by the postulated ejection of one of the hot leg thermowells is considered next.

If ejected, these thermowells with impact floors, walls, pipe supports or other steel structures which will not be affected adversely by impact.

Therefore, since no vital components which could sustain damage by impact are in the direct path of an ejected thermowell, the consequences of an ejected thermowell are bounded by the current small break LOCA analyses.

The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is,ot created.

The proposed changes will be performed in a mann=r consistent with applicable standards, preserve the existing design bases, and will not adversely impact the qualification of any plant systems.

This will preclude adverse control and protection system interactions.

The design installation and inspection of the'new equipment will be done in accordance with ASME Boiler and 4

s'

Pressure Vessel Code criteria.

By adherence to industry standards, the pressure boundary integrity will be preserved.

Hence, the possibility of a different type of accident than any evaluated in the safety analysis will not be created.

There will be no significant reduction in the margin of safety, as defined in the bases of any technical specification, since the Improved Thermal Design Procedure (ITDP) analyses remain bounding.

The applicable margins of safety are defined in Bases Sections 2.1.1 and 2.1.2.

Bases Section 2.1.1 states that the minimum value of the Departure from Nucleate Boiling Ratio (DNBR)-

during steady state operation, normal operational transients, and t<ticipated transients is limited to 1.17.

This value curresponds to a 95 percent probability at a 95 percent confidence level that Departure from Nucleate Boiling (DNB) will not occur.

The restrictions of this fuel cladding integrity safety limit prevent overheating of the fuel and possible cladding perforation which would result in the' release of fission products to the coolant.

The minimum DNBR reported in the accident analyses will be unaffected by the proposed change.

Bases Section 2.1.2 states that the Safety Limit on maximum RCS pressure is 2735 psig.

This Safety Limit protects the integrity of the RCS from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The maximum RCS pressure reported in the accident analyses is unaffected by the proposed change.

The proposed change will not result in a decrease of these margins of safety.

As discussed earlier, the response time and setpoints of the SG low-low level trip functions and the OPDT and OTDT reactor trip functions will remain within the assumptions used in the safety analyses.

As such, the analysis of the events which credit these functions will remain as presented in the FSAR, WCAP-10961-P, and WCAP-11883.

Consequently, the margins of safety between the Fuel Cladding Safety Limit (i.e., DNBR) and RCS pressure boundary Safety Limit and the actual failure of these barriers will not be reduced.

The setpoint calculations incorporate an uncertainty to account for the difference between the actual hot leg temperature and the measured hot leg temperature caused by the incomplete mixing of coolant leaving regions of the reactor core at different temperatures.

This uncertainty is made up of two parts -- a temperature streaming uncertainty and a scoop mixing bias.

The temperature streaming uncertainty is unchanged from the previous calculations as discussed in Section 2.0 of Attachment 6.

The scoop mixing bias is based upon an analysis of the effect of the modified hot leg scoop-thermowell design to be employed at Callaway.

The scoop mixing bias used in the setpoint calculations, in conjunction with a revised Delta-T gain (used to convert temperature in degrees F to percent Delta-T span) has been established as requiring only minor Technical Specification changes.._______

[

To determine the impact of the removal of the RTD Dypass piping and manifolds on the Callaway temperature-related control and protection functions, Union Electric performed instrument uncertainty calculations which utilized the latest available information on plant installed instrumentation and the Combustion Engineering scoop /thermowell design as well as previously approved setpoint methodology (revised slightly to reflect the additional RTDs and electronics).

As a direct result of this work, it can be concluded that the Rod Control system will operate within assumed tolerances, and the temperature-related protection functions, i.e.,

Overtemperature Delta-T and Overpower Delta-T reactor trips and Delta-T Trip Time Delays for SG Low-Low Level trips, will maintain their current Technical Specification Nominal Trip Setpoints.

Changes to the Technical Specification Z, S, and Allowable Values for OPDT and TTD Delta-T Power 1 and Power 2 will be necessary.

Based on these setpoint studies, the OTDT, OPDT, and SG Low-Low Level TTD instrument loops see reductions in the theoretical sensor and rack calibration uncertainty for each, primarily due to the increase in the number of RTDs and R/E converters used for the measurement of T-Hot.

The uncertainties used for the instrumentation remain. specific to the type and manufacturer of the hardware and are not a function of the presence, or absence, of the RTD Bypass piping.

The only uncertainties tnat change as a direct result of the removal of the RTD Bypass piping are the T-Hot streaming values.

It was determined that the CE scoop /thermewell design results in a small temperature measurement bias.

However, net reductions are seen in the instrument uncertainties which offset this bias and the revised Delta-T gain to the extent that only minor Technical Specification changes are required.

In conclusion, the removal and replacement of the existing narrow range RTDs and elimination of the bypass lines does not involve an unreviewed safety question.

Neither the probability of occurrence nor the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is created.

The possibility for en accident or malfunction of a different type than any evatunted previously in the FSAR is not created.

There will be no significant reduction in the margin of Ecfety as defined in the basis of any technical specification.

Therefore, the proposed revisions do not adversely affect or endanger the health or safety of the general public or involve a significant safety hazard, 1

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t SIGNIFICANT_ HAZARDS EVALUATIOlj i

This amendment application includes revisions to Technical

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Specification Tables 2-2-1, 3.3-4, and 4.3-1 to accommodate the proposed replacement of the current RTD bypass system'with an RTD/thermowell system incorporated directly into the hot and cold j

legs'of the reactor coolant system.

j The proposed change does not involve a significant hazards l

consideration because operation of Callaway Plant in accordance with this change would not:

(1)

Involve a significant increase in the probability or consequences of an accident previously evaluated.

The probability of.previously' analyzed accidents is discussed first.

The proposed change in RTD/thermowell response time will not increase the probability of such an accident because the response time is not a factor in the initiation of a previously evaluated accident.

The events of interest are those initiated by a failure of those components i

affected by the proposed change.

There are four such events:

a)

FSAR Section 15.4.2, Uncontrolled RCCA Bank' Withdrawal at Power

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b)

FSAR Section 15.1.3, Excessive Increase in Secondary Steam Flow c)

FSAR Section 15.1.4, Inadvertent Opening of a Steam Generator Relief or Safety Valve d)

FSAR Section 15.6.5, Small Break Loss of Coolant l

Accident (SBLOCA)

The Uncontrolled RCCA Bank Withdrawal event is an ANS Condition II-(moderate frequency) event potentially initiated by a failure of the rod control system.

Excessive Steam Flow'and Inadvertent.

Steam Generator Depressurization events are also Condition II events.

They are potentially initiated by a failure of the steam dump control system.

The input to the rod cont tl system and steam dump control. system from the replacement uxDs will be equivalent to that currently provided by the existing RTDs.

The proposed modification will be done in a manner consistent with the plant design bases.

As such, there will be_no degradation in l

the performance of or increase in-the number.of challenges to safety systems assumed to function in the accident analysis.

Furthermore, there will be no increase in the probability of 1

failure of or degradation of the performance of the systems designed to reduce the number of challenges to safety systems.

Hence, ther first three events will remain Condition II events.

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The SBLOCA is an ANS Condition III (infrequent) event.

It could be initiated by the highly unlikely ejection of a thermowell or the failure of one of the caps that will cover a

one of the existing cross over leg penetrations.

The i

scoops, cross over leg buttweld caps, and thermowells will be analyzed to the ASME Boiler and Preseure Vessel Code,Section III, Class 1 and installed in accordance with the requirements of Section XI of this Code.

As such, the RCS pressure boundary will not be degraded.- The SBLOCA will thus remain a Condition III event.

Additionally, approximately 400 feet of small diameter piping and the associated valves will be removed from the primary system' 1

pressure boundary, eliminating the possibility of a SBLOCA from these locations.

Hence,:there will be no significant increase in the probability of occurrence of an accident previously evaluated in the FSAR.

3 There will be no increase in the consequences of a previously evaluated accident.

In assessing the li.' pact on the consequences of a previously evaluated accident, there are six non-LOCA analyses of interest; a)

FSAR Section 15.2.3, Turbine Trip-b)

FSAR Section 15.4.2, Uncontrolled RCCA Bank Withdrawal at Power c)

FSAR Section 15.4.3, RCCA Misoperation d)

FSAR Section 15.4.0, CVCS Malfunction that Results in a Decrease in the Boron Concentration of the

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Reactor Coolant i

e)

FSAR Section 15.6.1, Inadvertent Opening of a i

Pressurizer Safety or Relief Valve f)

WCAP-lO961-P, Steamline Break Mass / Energy Releases for Equipment Environmental-Qualification Outside Containment These events are of interest because the OTDT trip is the primary trip assumed in the analyses, with a total response time of 8 seconds (6 seconds of first order lags and 2 seconds of pure delay).

No OPDT trips are assumed in the l

analyses.

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The OTDT, OPDT, and steam generator low-low level trips will i

continue to function in a manner consistent with the existing' analyses assumptions for these events.

The total response time of the proposed system will be less than that' j

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i for the current system.

Further, the elimination of the RTD bypass system will not affect the LOCA analyses input and the results of these analyses will be unaffected.

Therefore, the plant design changes due to the RTD bypass elimination are acceptable from a LOCA analysis standpoint without requiring any reanalysis.

Hence, there will be no increase in the consequences of previously evaluated accidents.

(2)

Create the possibility of a new or different kind of accident from any previously analyzed.

The proposed change will be performed in a. manner consistent with applicable standards, preserve the existing design bases, and will not adversely impact the qualification of any plant systems.

This will preclude adverse-control / protection system interactions.

The design, installation, and inspection of the new equipment will be done in accordance with ASME Boiler and Pressure Vessel Code criteria.

By adherence to industry standards, the reactor coolant pressure boundary integrity will be preserved.

As.such, the possibility of a new or different kind of accident is not created.

(3)

Involve a significant reduction in a margin of safety.

The applicable margins of safety are defined in Technical Specification Bases Sections 2.1.1 and 2.1.2.

Bases Section 2.1.1 states that the minimum value of the Departure from Nucleate Boiling Ratio (DNBR) during steady state operation, normal operational transients, and anticipated transients is limited to 1.17.

This value corresponds to a 95 percent probability at a 95 percent confidence level that Departure from Nucleate Boiling (DNB) will not occur.

The restrictions of this fuel cladding integrity safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the coolant.

The proposed change will not result in a decrease in the minimum DNBR reported in the FSAR accident analyses.

Bases Section 2.1.2 states that the Safety Limit on maximum RCS pressure is 2735 psig.

This Safety Limit protects the integrity of the RCS from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The proposed change will not result in an increase in the maximum RCS pressure reported in the FSAR accident analyses.

The proposed changes to the Callaway Technical Specifications are similar to changes approved at Byron Station Units 1 and 2 and at Salem Units 1 and.2.

As discussed earlier, the proposed change does not involve a significant increase in the probability or o

consequences of an accident previously evaluated or create the v

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< possibility of a new or'different kind of. accident from any j

.previously evaluated.:

It'does not result in a significant reduction in any associated safety limit or-limiting condition 1

of operation.

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Therefore, based on the above considerations, it has been i

' determined that the proposed change does not. involve a

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significant hazards consideration.

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