ML20151B813

Forwards Addl Info on 871109 Application for Amend to License DPR-6,changing Tech Specs to Reflect Features & Terminology Used W/New Power Range Monitoring Nuclear Instrumentation

ML20151B813
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	03/31/1988
From:	Berry K CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
NUDOCS 8804110288
Download: ML20151B813 (16)

Text

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Nuclear Lkeming M M N5 PROGREKE General Offees: 1945 West Pernell Roed. Jackson, MI 492ol e (517) 7881636 March 31, 1988 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT -

POWER RANGE MONITORING INSTRUMENTATION TECHNICAL SPECIFICATION CHANGE REQUEST ADDITI(;NAL INFORMATION Consumers Power Company letter dated November 9, 1987 submitted for NRC review and approval proposed technical specification changes to reflect the features 9nd tereinology used with new power range monitoring nuclear instrun9ntation.

The nee 1- "rumentation facility change is planned to be installed during the 1988 refueling outage which is scheduled to begin on April 8, 1988.

As stated in the Technical Specification Change Request submittal, the Big Rock Point Plant Review Committee and the Plant Safety Engineering Group under che cognizance of the Nuclear Safety Board had reviewed the facility change package in accordance with the requirements of 10CFR50.59 and Plant Technical Specifi-cations and determined the modification to not be an unreviewed safety question.

Subsequent to these reviews, the Nuclear Safety Board obtained additional information and has raised issues which has lead them to conclude the modifi-cation potentially may be an unreviewed safety question.

The Nuclear Safety Board conclusion is based on three considerations. The first is a reduction in the number of detectors from five to three may be a reduction in the margin of safety even through the reliability of the new instrumentation is increased and the fact that the new instrumentation incor-porates the functional requirements of the current intermediate end power range channels as described in current Plant Technical Specifications. The second is the new digital instrumentation may be susceptible to failures of a uifferent type than the previous analog instrumentation even though the design incorporates EMI filtering and a self-test system which continuously monitors the instrumen-tation for failures. The third involves a sensitivity analysis for the turbine trip without bypass transient in which credit for a more negative void co-efficient was taken to offset a slight increase in the scram delay time associated with the new instrumentation. Even though these issues have been raised, the Nuclear Safety Board concluded there is not a significant hazards consideration and the plant will continue to be operated safely with the new ,

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OCO388-0087-NLO2 8804110288 880331 PDR ADOCK 05000155

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2 instrumentation installed. These Nuclear Safety Board issues were provided to the Big Rock Point Plant Manager for resolution.

In response to the Nuclear Safety Board issues, the modification safety evaluation was revised to more specifically address the concerns and it was reviewed again by the Plant Review Committee. The Plant Review Connittee again concluded that the modification did not constitute an unreviewed safety question. The Nuclear Safety Board was satisfied with the Plant Review Committee resolution of their issues, but disagreed with the Plant Review Committee's conclusion concerning an unreviewed safety question.

Generally, the Nuclear Safety Board concerns are rooted in whether the appro-priate information has been provided to the NRC so that an appropriate level and depth of technical review can be accomplished by the NRC. This stems from the fact that our change request for the technical specifications associated with the new instrumentation categorized the changes as administrative in nature. It could be viewed that because of this classification, the NRC would treat the proposed changes as purely administrative and not perform a technical review. This could lead to an NRC decision based on incomplete information.

The details of the power range monitoring instrumentation modification have been provided to the NRC as supporting information to our November 9, 1987 Technical Specification Change Request. Also, it is our understanding that because a technical specification change is required, the level and depth of NRC review is not significantly different whether or not an unreviewed safety question is involved. It is our belief that the NRC has been provided with all of the information needed and that a thorough technical evaluation of our proposed changes is being performed.

The NRC reviewers have requested that we provide them a copy of our safety evaluation for the facility change to allow completion of their review of our proposed technical specification changes. As requested, a copy of the safety evaluation is attached. Resolution of tl.e Nuclear Safety Board issues has caused some delay in the submittal. We expect NRC review and approval can still be completed by the startup from our upcoming refueling outage. For your information, the outage is scheduled to begin April 8, 1988 and be conpleted on June 6, 1988. The power range monitoring, instrumentation modifi-cation is scheduled to begin April 18, 1988 and take 33 days to complete. If NRC review cannot be completed in this time frame, we would appreciate early notification so we do not place the plant in a condition in which it cannot be operated.

CW W g Kenneth W Berry Director, Nuclear Licensing CC Administrator, Region III, NRC NRC Resident Inspector - Big Rock Point Attachment OC0388-0087-NLO2

v e ,

I ATTACHMENT Consumers Power Company Big Rock Point Plant Docket 50-155 POWER RANGE MONITORING INSTRUMENTATION MODIFICATION SAFETY EVALUATION March 31, 1988 13 Pages OC0388-0087-N102

Rev 4

, Pago 1 ef 11 NUCLEAR OPERATIONS DF.PARTMENT BIG ROCK POINi PLANT SAFETY EVALUATION Item To Be Evaluated: Modification to Neutron Monitoring Instrumentation Item Identification: No. FC-599A I. 10 CFR 50.59 DETERMINATION

1. Is the item safety-related, or can it affect another safety-related item? List af fected item (s). _X_ Yes No

a. RT-RH01A and -B; Log-N/ Period Amplifiers (2)

b. RT-RH02A and -B; Log-N/ Period Amplifier Compensated Ion chambers (2)

c. RI-RH03A and -B; Remote Log-N Meters (2)

d. RI-RR04A and -B; Remote Period Meters (2)

e. RI-RH05A and -B; Dual High Voltage Power Supplies (2)

f. RT-RIO3A, -B and -C; Remote Flux Level Meters (3)

g. RT-TIO4A, -B and -C; Dual High Voltage Power Supplies (3)

h. RT RIO9A, -B and -C; Picoammeters (3)

i. RS-RIl0A, -B and -C; Picoammeter Range Switches (3)

2. Is the item changed from its description in the FLUt/FHSR? List affected section(s). X Yes No

a. THSR Sections 7. 5. 6, 7. 5. 7, 7. 6.1. 2 and Table 7.2

b. Docketed information - SEP Topic VI-10.A and

. letter dated July 2, 1985

3. Does the ites involve a test or experiment not previously described in FSAR/FHSR? Yes X No

4. Does the item require a change in Technical Specifications? List affected section(s). _X_ Yes No Technical Specifications 6.1. 2, 6.1. 2. 2, 6.1. 2. 3, 6.1.2. 5, 6.1.5. c , d and e , 6. 2.1.b , 6. 3.1, 7. 3. 5.e , Figure 5.1 M10388-3111A-BT01

Rev 4

. Page 2 of 11 II. 10 CFR 50.59 EVALUATION

1. Is the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluaced in the Safety Analysis Report increased? Yes _X_ No Basis:

The replacement of the existing instrumentation is based on the need to have a stable, reliable neutron monitoring system to provide monitoring of the neutron flux characteristics associated with reactor operation. The DC Wide Range Monitor (DCWRM), manufactured by General Electric Company, is a member of the Nuclear Measurement Analysis and Control (NUMAC) family of microcomputer based instruments and has been selected to replace the Log-N/ Period (intermediate range) and Picoammeters (power range) channels at Big Rock Point Plant.

The DCWRM is designed to monitor the intermediate and power ranges of a nuclear reactor by measuring the input current from a neutron sensitive compensated ion chamber, calculating power level, reactor period, power rate and performing trip and alara functions based on these calculations.

Each DCVRM displays percent of full power and reactor period or power rate, both digitally and on a logarithmic bar graph. Analog and digital outputs are provided for remote meters and/or recorders.

The DCVRMs are capable of performing the following functions:

Measuring input current from a compensated ion chamber end performing specified power, period and rate-of-change calculations.

Providing high voltage detector excitation (polarization) power.

Providing power / period / power rate output data to remote seters and/or recorders, either in analog form nr over an RS-232 digital data link.

Providing alarm, trip and control rod block / permissive signals.

• Automated calibration

• Automatic self testing and alarm, with self-test status displays on demand.

6 Providing security against tampering (by password and keylock control).

Utilization of the DCVRMs permits replacement of the two (2) Log-N/ Period amplifiers and their associated dual high voltage power supplies, remote metering and detectors and the three (3) Picoammeters and their associated range switches, dual high voltage power supplies and remote metering. The existing power range (Picoammeter) detectors, compensated ion chambers identical to those used with the Log N/ Period amplifiers, will be retained for use with the DCWRMs.

MI O383-0111 A-BT01

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. Page 3 of 11 Even though two (2) channels (Log-N/ Period amplifiers) of neutron monitoring system instrumentation are being eliminated, the DCVRM combines the coverage of the flux range to be monitored, provides relay outputs for control rod block / permissive signals in accordance with plant requirements and provides output relay trip signals to the Reactor Protection System which equal or exceed the existing trip signals.

Elimination of the Log-N/ Period amplifiers is made possible because of the wide range of the DCWRM (1 x 10 7 1 to 150% power) and its ability to perform all of the required monitoring and tripping functions of the Log-N/ Period amplifiers. Visibility to the intermediate range of neutron flux monitoring is improved, as three (3) DC%Tds will be provided (previously two channels).

The intermediate range portion of the DC%RM will provide neutron flux level coverage from 1 x 10 7 to 1% power. Three intermediate channels, providing short period protection in a 2 out of 3 logic input network to the Reactor Protection System, will improve plant reliability by reducing challenges from spurious trips associated with the former 1 out of 2 logic. An analysis of the unavailabflity to perform the trip function for the 2 out of 3 lo*,ic versus the 1 out of 2 logic for the specific Big Rock Point installation has been performed and concluded that the 2 out of 3 logic has a better (reduced) percentage of unavaila-bility. This reduction is based on the fact that with the DCVRM instrument, regular periodic testing can be performed.

A study of the failure history associated with the Log-N/ Period amplifiers has been performed. The failures include only those attributable to the Log-N/ Period unit, not the detectors nor the high voltage power suppl'.es which apply polarizing voltages to the detectors. These iter, are not included in the failure history because 1) the existing detectors will be used with the NUMAC system and 2) the existing power supplies are not dedicated to the intermediate channels and c n be used in both the intermediate .

range channels and the povar range channels. Failure histories are kept by equipeent scrit.1 number, not channel location. The absence of the hish voltage power supply failures will obviously tend to bias the following unavailability analysis in favor of the 1 out of 2 trip logic presently in use.

Since 1963, there have been 86 documented failures associated with the Log-N/ Period amplifiers. These failures have been attributed to three (3) units, two of which are always in service and one spare. The Log-N/ Period amplifiers are not repaired in place but can normally be exchanged in one hour; therefore the mean time to repair (MITR) on these units is considered to be one hour.

The total operating time of the Log-N/ Period amplifiers for the historical time period is equal to:

2 (25 yrs) (365 days /yr) (24 hrs / day) = 438,000 brs.

The mean time between failure (if7BF) is therefore 438,000 t 86 = 5093 hours0.0589 days <br />1.415 hours <br />0.00842 weeks <br />0.00194 months <br />. With a MTBF of 5093 hours0.0589 days <br />1.415 hours <br />0.00842 weeks <br />0.00194 months <br />, the failure rate is equal to 1.72 failures per year.

MIO388 0111A-BT01

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. Page 4 of 11 In comparison, the huAC-DChM Performance Specifications (23A5052) indicates that the MT1T for the overall instrument (including the operator display and front panel display) is 3100 hours0.0359 days <br />0.861 hours <br />0.00513 weeks <br />0.00118 months <br />. F_xcluding the display features as they have no impact on the instrument's ability to perform its safety function yields a PrTBF of 17,360 hours0.00417 days <br />0.1 hours <br />5.952381e-4 weeks <br />1.3698e-4 months <br />. Utilizing a MTBF of 17,360, the failure rate is equal to 0.505 failures per year.

Unavailability as a function of logic configuration and testing schedule is defined as follows : 1 For a 1 of 2 system: U = 1/3 (AO)2(goo)

For a 2 of 3 system: U= (AO)2(goo) k'here U = percent unavailability per year A = failure rate per year 0 = testing interval per year (simultaneous)

Using the failure rates established in the previous paragraphs, the unavailability for each logic configuration based on a test interval of once per year becomes:

1/2 system: U = 1/3 (1.72

• 1)2(100) = 98.6%

2/3 system: U= (0.505

• 1)2(100) = 25.5%

This illustrates that the unavailability is much better for the 2/3 system than the 1/2 system presently in use at Big Rock Point, based on observed or predicted failure rates and a testing interval of once per year. The testing interval of once per year is proper for this specific case because the existing Log-N/ Period a:rplifiers cannot be functionally tested above N1% power (ie, when the picoansneter range switches are placed in the 4% position). Normally, the trip functions of the Log-N/ Period amplifiers are only checked during reactor startup. .

The N12iAC-DC%M instruments have the capability to perform on-line testing (and actual operation of the trip relays) during power operation. The trip levels of power trip, period trip and rate-of-power change trip can be tested with the NLHAC-DChM in "INOP" (placing the keyswitch in "INOP" places the instrument in a downscale trip positico, reverting the logic to a 1/2 on the other channels).

At the present time, the picoacceter trips are tested on a monthly basis. Utilizing this same test frequency on the NLHAC-DCW, the unavailability becomes:

2/3 system: 0 = (0.505

• 1/12)2(100) = 0.177%

Further, the NUMAC-DCM is equipped with a self-test feature which provides a complete check of firmware and hardware. This test f requency has not been firmly established as it is dependent upon instrument response time (with an instrument response time of 310 msee the test frequency was twice per minute; with a reduced response time , ie,1100 msec, the test frequency will be greater MIO388-0111A-BT01

R:v 4 Pass 5 of 11 but not exceed a frequency of once every five minutes). If credit is given for self-test, the unavailability becomes:

2/3 system: U = (0.505

• 1/105120)2(100) = 2.31E-09%

Based on the above calculations, it can be demonstrated that the reliability of the circuitry in the trip mode for the application of the NUMAC-DCVRM instruments at Big Rock Point Plant has increased and the probability of a malfunction has decreased with installation of the 2 out of 3 period trip.

With the inclusion of the derivative circuitry (ie, reactor period) in the NUMAC instrumentation,'the benefit of obtaining greater reliability of detecting rising neutron fluz is achieved. Three neutron detectors of identical sensitivity to those used for the intermediate channels being removed, together with a closely coupled Big Rock Point nuclear core increase the probability of seeing a short period condition. The margin of safety for detection of rising neutron flux is increased by the utilization of three detection mechanisms spaced radially at 120' around the reactor core versus the two detector channels radially spaced at 180' presently in use.

Therefore, the monitoring and protection capabilities provided by the original intermediate channels has been maintained and the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report is not increased.

• Replacement of the power range (Picoameters and associated range switches) instrumentation by the DCVRM will provide the plant with a more reliable power range neutron monitoring system. The power range portion of the DCVRM will cover neutron flux levels from 1% to 150%.

Reliability improvements are realized by the elimination of the pitoammeter range switches which have the potential for spurious noise when changing scales; this results in unnecessary challenges to the Reactor Protection System. Further, the DCWRM has the capability to provide all of the auxiliary functions (eg, rod block / permissive interlocks, refuel setback, etc) associated with the range switch.

Alarms and trip functions that are required for the monitoring of neutron flux levels and tripping of the Reae.or Protection System ' ill be maintained.

Automat.ic reset of the "HI-HI TRIP" (eg, Reactor Protection System input) will be utilized in place of the manual reset function presently associated with the range switches. The automatic reset vill occur af ter a trip latch time of 2100 asec to ensure that the Reactor Protection System has had sufficient time to receive the trip signal.

Automatic reset however, will reduce spurious challenges and hence improve plant reliability without reducing trip capability. The 2100 nsec time interval has been analyzed and is deemed appropriate for the Neutron Monitoring System / Reactor Protection System response inter-action as discussed below:

M10!$8-0111 A-BT01

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. Page 6 of 11 The original picosameters and Log-N/ Period amplifiers utilize blocking oscillator type trip units which require only 1.5 milliseconds to drop to a low (0-volt) state when the trip level is reached.

Even though the picosameter utilizes a manual reset (ie, trip latch) in its trip circuit, the Log-N/ Period amplifier trip resets automatically; the delay in reset based on the hysteresis associated with the trip unit. The ability of the Reactor Protection System to sense the Log-N/ Period trip has never been questioned, even though 1.5 milliseconds is a very short time period.

The NUMAC equipment trip relays are deenergized via the Input /

Dutput (I/0) module. The I/O module trip levels have discrete hysteresis levels; ie, for each trip level there is a reset level vtich must be reached before a trip relay will be energized. For example, on the 120% power trip, the pcver must decrease to 118%

before the trip relay will be energized. Other trips have similar, or greater hysteresis levels. In addition to this, the trip relays have timing values of 10 milliseconds for dropout and 13 milliseconds for pull-in. Therefore, the trip delay time associated with the NUMAC equipment equals or exceeds the delay time associated with the existing equipment.

However, to provide additional conservatism in the trip delay time, the trip relays in the NUMAC equipment vill be programmed for an automatic minimum "latch" time on trips of 100 milliseconds. This vill ensure that the Reactor Protection System "sees" trip signals that may not arrive simultaneously on all three channels. The 100 millisecond time interval was chosen, as this amount of time will be equal to or greater than the response time of the Reactor Protection System.

The response time of the DCVRM is not as fast as the original power range monitors. The response time of the original power range is specified as $21.5 asec; the response time of the DCVRM is $110 msec.

This change in respon .e time has been analyzed and it has been determined that an increase in response time up to 125 msee is permissible as discussed below:

The response time for the KUMAC-DCVRM is 5100 msec. In this case, due to the reconfiguration of the Reactor Protection System inputs to utilize the trip relays associated with NUMAC-DCWRM, the contact operating time of the trip relays must be included. The drop-out time is 10 msee making the total response time $100 msee plus 10 asec or $110 msec.

Recently, an analysis utilizing the sensitivity study previously considered on the Big Rock Point Accident and Transient Analysis was performed to verify the amount of time available for instrument response prior to Reactor Protection System trip. In previous analyses, this time (521.5) msec) was considered insignificant compared to the Reactor Protection System delay time of 100 msee plus the scram valve operating time of 375 msec (ie,100 msee plus 375 msee or a total delay time of 475 asec).

M10388-0111A BT01

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. Pag 2 7 ef 11 A minor change to the void coefficient of 101 (ie, -0.0209047 Ak/k;1001 voids) improved the change in MCPR (Minimus Critical Power Ratio) for the most limiting accident by decreasing the amount of positive reactivity insertion associatcd with this transient. This change provides an additional delay of 125 maec for monitor response time 2 resulting in a total delay time of 600 msec. The present fuel cycle (and previous cycles) have been well within this new void coefficient of -0.188142 Sk/k/100% voids (see below).

Cycle BOC = Void (ak/k/100 Voids)_

17 -0.1453 18 -0.15797 19 -0.17701 20 -0.16216 21 -0.16295 22 -0.15576 23 -0.16563 (preliminary)

A review of the Control Rod Drop Accident Analysis was performed to confirm that the additional time delay associated with the NUMAC-DChEM response would not have a significant impact on the total deposited enthalpy in the fuel. The reactivity feedback due to Doppler essentially terminated the event prior to scram insertion. The analysis utilized a delay time of 375 msec. The additional delay time of 225 msec (eg, 600 msee - 375 msec) results in an increase of 53 cal /gs, which is considered negligible .8 The present fuel cycle (and previous cycles) have been well below the 280 cal /gm limit set in Exxon Report KN-NF 78-51 (see below).

Total Deposited Enthalpy (cal /gn)

. Cycle 375 msee Delay 600 usee Delay 20 157 160 21 204 207 22 195 198 23 203 206 (preliminary)

A second method of protection has been included in the DCVRM to provide closer coverage of flux changes during power ascent from 1% to 100%

power. This additioosi layer of protection was added to replace range-related trips associated with the picoammeter range switches even though no credit was ever taken nor assumptions made which relies on range related trips to mitigate the consequences of any analyzed accident or transient included in the Big Rock Point Accident and Transient Analysis.

The primary function of the range switches was to ensure usable indication to the operator over the entire range of flux to be monitored and was not inconsistent with the technology available at the time the picoammeters were built. The obvious added feature to such construction (ie, switching of the measured ranges in 3.16% steps) permitted the trip set point of 96% of scale to be held closer to the actual measured MIO388-0111A-BT01 m __

R:v 4 Paa2 8 cf 11 power level, irrespective of which step the range switch was on. This feature provided additional protection by keeping the trip point closer to tetual power level. It should be noted, however, that the range switches were placed Um the 125% powet position when the reactor reached a power level of approximately 30%.

Early in conceptual design associated with the replacement of the picoammeters with the VUMAC-DCWRM, it was realized that an additional circuit function, ie, rate of change of power could be utilized to incorporate trips or rod permissive signals if a rate (or rates) were exceeded. This feature is included in the NUMAC DCVRM design to compensate for the loss of range switch coverage as described in the previous paragraph.

With the rate-of-change inccrporated into the NUMAC-DCWRM logic, additional protection is available in the order of rod permissive signals and reactor trip capability if certain rates of power level change are exceeded. This provides greater protection than what was previously available. This protection incorporates the principal core operation limitations as defined in Plant Technical Specifications 5.2.1 and includes the following:

Reactor Power  % Power Rod Block B1-HI TRIP 5120 MWt 1% to 50% 237.5 MWt/ min 150 MWt/ min 2120 to $200 MWt 50% to 83.3% 215.0 MWt/ min 120 MWt/ min 3200 to $240 MWt Above 83.3% 210.0 MWt/ min None This rate-of-change protection does not replace any required HI-HI TRIPS on reactor power level but is an additional layer of protection designed to provide scre conservatism in design and limit the reactor power ascent in accordance with the Plant Technical Specifications by automatic action. The end result of inclusion of such a protective circuit does not invalidate the power channel trip circuitry; rather, the trip circuitry is enhanced and does not result in the~ possibility of a different type of salfunction.

Therefore, it can be concluded that the monitoring and protection capabilities provided by the original power range monitoring system has been maintained and the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report is not increased.

2. Is the possibility of an accident or a different type than any evaluated previously in the Safety Analysis Report created? ___ Tes X No Basis:

Utilization of the DCVkM vill result in the elimination of two intermediate range monitors; the DCVRM is capable of providing this neutron flux monitor-ing function and short period protection from 1 x 10 to 1% pover.

7 MIO388-0111 A-BT01

l Rev 4

- Page 9 ef 11 The existing Plant Technical Specifications, Section 6.1.5(d) define that period protection will be provided up to approximately 5% rated power. lu reality, bypass of period protection has never occurred at greater than approximately 1% power at Big Rock Point. The period bypass occurs when the operator places the range switch in the 4% power range. This usually occurs at approximately 65% of the 1.25% power range or at approximately 0.8% power. Even though the existing Plant Technical Specifications (Section 6.1.5.d) state "For reactor operation above approximately 5% of rated power, the logarithmic neutron flux level information and period scram protection are not required (see Section 6.1.2)", original plant design as discussed in the Final Hazards Summary Report (FHSR), Sections 7.6.1.2 and Table 7.2 clearly identify the bypass of period trip circuitry occurring at 1% power. Further, in Consurer Power Company's response to SEP Topic VI-10. 4A , the following comment is made in Section 3.1.5.5.

"Period scram signals are bypassed when the range switch of any two power range piccammeters are in the 4% full-scale (or higher) power range positions (see the ausiliary switch S4 development table in Attachment 3 and the period bypass circuitry which energizes relays IK6 and 2K6 in Attachment 1). This bypass feature is used to avoid spurious scrass by fluctuating period measurements caused by steam voids when boiling occurs in the core."

Apparently, confusion regarding the actual point at which bypassing of the period protection occurs has existed since the inception of the Big Rock Point Plant Technical Specifications due to the f act that the bypass takes place when the range switch is placed in the 4% power position, not at 4% power. The proposed changes to the Technical Specifications will accomplish this and fix a firm point (ie,1% power) at which the bypass occurs in the DCVRM.

The period trip circuitry will be changed from a 1 out of 2 logic to a 2 out of 3 logic. In addition to providing greater plant reliability, .

analysis has shown that for this particular installation and considering past f ailure history and predicted failure rate of the DCVRM, the percent of unavailability is reduced through use of the DCVRM.

Utilizetion of the DCVRM will result in replacement of the existing power range monitors and their associated range switches; the DCVRM is capable of providing this neutron flux monitoring function and high flux trip capability ftem 1% to 150 power. In addition, the DCVRM can fulfill all of the auxiliary functions for control rod block / permissive signals, refuel setback, etc, associated with the auxiliary functions of the range switch.

The neutron monitoring inputs to the Reactor Protection System are shown in Attachment 1. Relay logie utilizing the 26 VDC derived from the Reactor Protection System logic power supply will replace the voltage level inputs to the Reactor Protection System for Neutron Monitoring Channels 1, 2 and

3. The logic configuration is straight forward and less complicated than the present upscale-downsci;e trip voltage logic associated with the existing Reactor Protection System inputs from the Neutron Monitoring System. Note that the period inputs (5-4 and 5-6) are hardwired to eliminate the period trip function in the proposed configuration.

1:0368-0111A-BT01

• ' Rsv 4 Page 10 of 11 The control rod permissive functions associated with the power range monitoring instrumentation are shown on Attachment 2. The notes provided on the attachment show the existing and proposed contact configurations.

Note that a new relay (4K1) will replace the existing contact configuration and be cont rolled by the rod block network. This relay is added to provide greater relay contact capability in the manual reactor control system and provide additional isolation between systems. Also, the suppression network 4Rl-4C1, included to protect the range switch contacts, is no longer required.

Therefore, it is concluded that the DCWRM will provide the necessary flux sonitoring capability from 1 x 10 %7 to 1502 power, provide the necessary tripping functions as inputs to the Reactor Protection System and provide the required interlock capability for the control rod drive (manual reactor control) system; the possibility of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report is not c rea t ed.

3. Is the margin of safety as defined in the basis for acy Technical Specification reduced? Yes X No Basis:

The DCVRM will fulfill the monitoring, tripping and interlock features as required in the basis for the appropriate Technical Specifications. Even though the response time of the DCVRM is somewhat slower, analysis has shown in Part 1 above, that the safety features associated with reactor tripping are not compromised. Therefore, it is concluded that the margin of safety as defined in the basis for any Technical Specification is not reduced.

References:

1. ANSI N41.4-1976/IEEE Std 352-1975, IEEE Guide for General Principles of Reliability Analysis of Nuclear Power Generating Station. Protection Systems.

2. 105; HGBazydlo, CPCo to GRBoss, CPCo Big Rock Point dated March 3,1988 entitled "JUSTIFICATION FOR OPERATION VITH INCREASED SCRAM TIME ARISING FROM NElTTRON MONITOR RESPONSE TIME".

3. Telecopy: O. Craig Brown, Advanced Nuclear Fuels Corp. to RESchrader, CPCc - Big Rock Point dated March 14, 1988.

4. Letter, RAVincent, CPCo to DMCrutchfield, USNRC, dated March 29, 1982 entitled "BIG ROCK POINT PLANT - SEP TOPIC VI-10. A, ELECTRICAL, INSTRUMENTATION AND CONTROL PORTIONS OF THE TESTING OF RTS AND ESF".

III. NRC NOTIFICATION

1. Should this be included in FSAR/FHSR update? X Yes No

2. Should item be included in Annual Report to NRC1 X Yes No MIO388-01HA-BT01

e Rev 4

. Page 11 et 11

3. Is prior NRC approval and an application for amendment to License required? X Yes No IV. APPROVALS Prepared by: I .* 3[ .( Il<t .

g2 'e** 2) Woutron Monitoring Channels 1. 2 sad 3 g

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1 4.rs.t a b>h ,l

- upscale-dewnscale cotacident trip is bypassed with all three power channel 7s4 I 1.Ca , ,,va.h.q g

J bb ' 'l l range switches to the 40 x 10 11 power y% y

~# position er below.

2,. E f I *' ** 8' ' #

3) Neutron Moottertog Char.nels 1. 2 and 3

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specale-downscale relaying is part of the l Bestron Monitoring System trip output logic.

C*d wr4 $ 9,- b W nra Md C '

4) Lt6 relay contacts are closed (ie, short period trip bypass) with two of three power channel raege evitches la the et power i : > h^ position (actual bypass operation occure at

1. 6 r ; _ m_ __ te l appreminately 60 to 70! of the 1.2SI power N eo s / aaa8 ; se i

' posittee or at 0.75 to 0.8731 power when the 8 *

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1.

Altstace Nkew. re te.t ne setAsce utv emee,we8,S u saa ,, > t 87sr .a FIGURE 1 PROPOSED CONFIGURATION

1. 6 "  ? Retest h.e s s ~ 4 ras ~l M)- Nh' '

!) Neutroe llen.iterits Chaamals 1. 2 ad 3

!f.M e

yh w DCVRM "I1-51 Trip" ecevre at 31201 povert

$10 second period when below 11 power or Q Q'p% fast power rate when above !! power (ie.

i n( a hb

'8 34M r f ree 11 to 501 power 450 wt/ minutes f ree

l 3 6)

- 501 to 83.31 power 420 W t/ minute).

lg M gp Contacts 1. 2 and 3 opea ea"El-81 Tttp" M ,

b w 2) Contacts D1. D2 and D3 opes apon toetrument down-ocale (1 a 10*i1 power), high voltage hy H .5 D

j' power espply f ailure (+ or - er both at t 101 et voltage setting of 3007 DC). key lock er .:ao evitch is "Inor" posittee er "Fatal Fault

• h ta NINAC DCVRM hardware /firmware. Ope ning h

C of these coatecto *eets up" the upscale-downecale trip logic ta provide reacter scran ee a "31-51 Taip" f rom another

8 ^ chamael.

___e mee eJf & 0  ;.

,es -

,,e......,t . u

o an.ame . w m.= e= =me,w e =e,rc w row FIGURE 2

~.

. v I ATTACHMENT 2 C0KTROL ROD PEPMISSIVE CIRCUITRY EXISTING CONFIGURATION isies, aw l) 1. 2. 3 . Neut roe Meetering Channele 1. 2

. 4ast and 3 downscate ingerlock bypass. Costacte

' closed at 60 a 10' & pcver ramte evitch

......... ............. 3 position and below perstettag control rod

. wit hd r awal .

l1

.*I e  :) 14. 2d. 3d . Neutron Monitories ;bannels 1.

'y g 4g

- Id p-

• 14 8 1 and 3 downscale interlock relay contacts.

'.1- 8 y ui kh Tt T Iy '

Contact closed with picosameter reading 441 r t r atas a= 7 c 2- (2 5 '*-

'7 ' 125 scatel 1.6 on the 40 ecale) perettting 6 .......... ...............J cutrol rW withdrawal.

3) The picosameter range evitch has nineteen RE FE RE *1 DtAw th4 ; (19) scales idenettled as folleve CPCS 07447 30733 sar.2 Switch ioso 40 a 10*8 125 a 10*8 (64.104 A #94 s a e) 40 10 ........ 40. 125. talp itst.

FIGURE 3 PROPOSED CONTIGURATION est ee t j  !) 1. 2. 3 - Bestron Itsettorias Channele 1 2 and 3 DCwel "El Tat?" at 4 10$ power.

% s 15 eeceed pertad (between 1 s 10*? and l.01 powe r) er f as t powe r rat e ( f rom 1.0 I

$ U' {

. 4 Ell to SOS power a 37.5 wt/minutel free 50 g o a to 83.31 power 415.0 Wt/staute t above i ---* ----y 3.31 power 310.0 wt/stante). Contacts uo t -I

-- I .. e opes os "El 711P" condittee to block costrel o I rod withdrawal.

{Q e

. 48tl e o

i 2

p

) e. 7 . sever.. Moottutas n .-la 6 ud StM count rate bTyese. Contacta closed og 6..1 1W g

" between and 4 s 10' e and 4.000 epe (* I 101 I egI power) bypassing downacale gg taterlocks and permitting control red 3-= 1 l withdrawal.

tu ag T ^ r $d i 3) 14. Id. 3d - Westres Moottorias Channele 1 I l 2 and 3 DCVRM dowmacale laterteck relays.

4El Contacto closed above I 10* M power ir g (prelistaary setting, say be raised to enevre power tlLannel operabtitty in (REFEti,tsC(

, m m y ,3 M,,, ild61m au erdance viih pl.et t u hai.a1 M ctftcatt u 6.1.S.c) peretttles I $.E.104 R19 4 Jahl) costrel red witherswal.

1 i

FIGURE 4