Text

Safety Evaluation Supporting Amend 71 to License NPF-62
	ML20035F646
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Site:	Clinton
Issue date:	04/09/1993
From:	; Office of Nuclear Reactor Regulation
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 71 TO FACILITY OPERATING LICENSE NO. NPF-62 ILLINDIS POWER COMPANY. ET AL.

CLINTON POWER STATION. UNIT NO. 1 DOCKET NO. 50-461

1.0 INTRODUCTION

In a letter dated September 20, 1991, Illinois Power Company (IP), the licensee, proposed changes to the Technical Specifications (T3) for the Clinton Power Station (CPS).

In response to questions from the staff, IP supplemented the proposed changes in letters dated August 17, 1992 and February 17, 1993. These supplements consisted of revisions / clarifications which did not change the staff's initial proposed no significant hazards consideration determination. The proposed TS changes incorporate reliability based improvements to instrumentation Action Statements and surveillance test intervals based on NRC approved Licensing Topical Reports (LTR) previously submitted by the Boiling Water Reactor Owners Group (BWROG), of which IP is a member.

This safety evaluation verifies that the TS changes proposed by the licensee and their associated justifications are bounded by the analyses provided in the approved topical reports and that any plant-specific issues are adequately addressed.

2.0 EVALUATION The licensee's proposal is divided into five separate packages: (1) Reactor Protection System (RPS); ? ) Emergency Core Cooling System (ECCS); (3) Control Rod Block; (4) Containment and Reactor Vessel Isolation Control System i

(CRVIS); and (5) Other Technical Specification Instrumentation. The staff evaluation is divided in the same manner below.

Staff approval of plant-specific TS changes was conditional, as described in the individual LTRs, upon satisfactory resolution of plant-specific conditions as described below.

(1)

Licensees must confirm the applicability of the generic analyses for the LTR to their facilities.

(2)

Demonstrate, through the use of actual instrument drift information from the equipment vendor or from plant-specific data, that the drift characteristics for instrumentation used in the plant are beunded by the assumptions used in the LTR when the functional test interval is extended from weekly or monthly to quarterly.

9304220125 930409 ADOCK0500g1 DR

(3)

Confirm that the differences between individual components of the RPS that perform the trip function installed in the facility are bounded by the generic analysis.

2.1 Reactor Protection System (RPS) 2.1.1 Introduction On January 24, 1988, the staff forwarded to the BWROG its safety evaluation report (SER) on the review of the following General Electric (GE) Topical Reports: (1) NEDC-30844, "BWR Owners' Group Response to NRC Generic Letter (GL) 83-28, " Required Actions Based on Generic Implications of Salem ATWS Events;" and, (2) NEDC-30851P, " Technical Specifications Improvement Analysis for BWR RPS."

NEDC-30844 forwarded the BWROG's technical basis that current BWR RPS surveillance test intervals (STIs) meet the recommendations of GL 83-28, Item 4.5.3.

NEDC-30851P utilized the base case results of NEDC-30844 and fault tree modeling to establish a basis for extending the current RPS STIs and allowed out-of-service times (A0T).

The staff's review of the above mentioned Topical Reports concluded that: (1) the failure data and methodology utilized to determine the initiating event frequency and RPS unavailability data were generally adequate for both the solid-state and relay designs; (2) the solid-state design is much less sensitive to the considerations raised in GL 83-28; (3) the extension of the A0T when two sensor channels are out-of-service from I hour to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is acceptable; and (4) the potential benefits associated with an increase in the STI for RPS instrumentation outweighs the very small increase in core melt frequency.

In addition, the staff concluded that each licensee must address instrument drift and verify that the LTR is directly applicable to the CPS in their application.

The staff's SER on the above Topical Reports, dated January 24, 1988, is directly applicable to BWR plants with solid-state RPSs.

CPS is the only BWR with a solid-state RPS; therefore, the staff's safety evaluation is currently applicable to this plant alone.

2.1.2 Technical Specifications The licensee proposed the following changes to the RPS TS which have generally i

been reviewed and approved in the staff's SERs on Topical Reports NEDC-30844 and NEDC-30851P, and in a letter dated July 26, 1991:

2.1.2.1 Allowed Out-of-service Times A. Technical Specification changes:

TS 3.3.1, Action a.2. - TS change increases the time to place one inoperable channel in the tripped condition from I hour to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months when two of the four channels required for any trip function become inoperable.

1 6

Footnote to TS 3.3.1 - A0T for surveillance is increased from 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

B. Potential " loss-of-function" issue for the RPS:

The staff expressed a concern regarding the proposed generic model TS change (ACTION "a") to the BWROG in a letter dated July 26, 1991. The staff indicated in the subject letter that model ACTION "a" could i

potentially be read to mean that loss of RPS functional capability is f

permitted for the 12-hour A0T. After discussing this concern with the BWROG and GE, the BWROG and GE agreed that the assumptions of the NEDC-30851P-A analyses did not support the explicit language in the model TS and that new TS wording would be submitted for staff review; furthermore, the BWROG emphasized that operators would not willingly allow this condition to exist. The staff indicated in this letter that they would review individual license applications to ensure the explicit language originally proposed as the model was corrected to prevent this interpretation of the TS ACTION "a".

The licensee states in their analysis that the potential " loss-of-function" issue is not applicable to the solid-state RPS. The plants utilizing a relay type RPS have a one-out-of-two-taken-twice logic.

With this type of logic, if both channels of two parameters monitored were to become inoperable, conditions could exist where a " loss-of-function" could occur. The CPS solid-state RPS logic is arranged in a two-out-of-four type logic scheme. With the proposed TS and the RPS logic scheme, CPS could not operate when any parameter is lost such that a " loss-of-function" could occur.

Therefore, the licensee has determined that the potent' ~ " loss-of-function" issue is not applicable to the solid-state RPS at (JS.

The staff agrees with the licensee's assertion.

t The licensee's analysis indicates that NEDC-30851P-A and the staff's associated SER, dated January 24, 1988, directly correspond to the CPS solid-state RPS design.

In addition, the licensee has adequately addressed the proposed TS concerns regarding the potential " loss-of-function" issue for the RPS.

Based on the above, the staff finds the licensee's proposed TS change acceptable.

l 2.1.2.2 Surveillance Test Intervals A. Technical Specification changes:

i The CHANNEL FUNCTIONAL TEST interval specified in TS Table 4.3.1.1-1,

" Reactor Protection System Instrumentation Surveillance Requirements,"

l is increased from weekly (W) or monthly (H) to quarterly (Q) for the t

following FUNCTIONAL UNITS:

2.b.

Average Power Range Monitor (APRM) Flow-Biased i

Simulated Thermal Power - High (W to Q),

)

_4_

2.c.

APRM Neutron Flux - High (W to Q),

2.d.

APRM Inoperative (W to Q),

3.

Reactor Vessel Steam Dome Pressure - High (M to Q),

4.

Reactor Vessel Water Level - Low, level 3 (M to Q),

5.

Reactor Vessel Water Level - High, Level 8 (M to Q),

6.

Main Steam Isolation Valve - Closure (M to Q),

7.

Main Steam Line Radiation - High (M to Q),

8.

Drywell Pressure - High (M to Q),

9.a.

Scram Discharge Volume Water Level - High Level Transmitter (M to Q),

10.

Turbine Stop Valve - Closure (M to Q),

11.

Turbine Control Valve Fast Closure Valve Trip System 011 Pressure - Low (M to Q),

13.

Manual Scram (M to Q)

Table 4.3.1.1.-1 Table Notation (g) - Analog trip module calib ation periodicity increases from 31 days to 92 days.

B. Instrument Drift The licensee indicated in its safety analysis that the currently approved RPS instrument setpoint calculations include the effects of instrument drift over surveillance intervals of 18 months for all instrument loops with the exception of the analog trip modules. -To address the drift issue with the RPS analog trip modules, IP indicated that it reviewed the results of monthly calibration checks performed over a 1-year period. As a result of this review, IP concluded that the quarterly drift is within present tolerances to allow an increase in STI for the RPS analog trip modules from 31 days to 92 days.

Records showing the actual setpoint calculations and supporting data should be retained on-site for possible future staff audit.

The licensee's analysis of actual surveillance data appears to bound both linear and non-linear drift.

In addition, the licensee indicated that plant procedures require operators to perform a CHANNEL CHECK at least once per every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for instruments with redundant channels.

Many instruments, including the RPS analog trip modules, have redundant channels; therefore, should drift begin to occur, the licensee has the capability to identify the problem early and take appropriate corrective action.

Based on the licensea's analysis of instrument drift, the verification of fault tree modeling data and LTR applicability, and the acceptability of the proposed changes, the staff concludes that sufficient analysis has been performed to support the proposed RPS STI_ increases indicated above.

2.1.2.3 Editorial Technical Specification Changes An editorial change deleting footnote "*** associated with Surveillance Requirement 4.3.1.2 is acceptable because it was applicable until the first refueling outage which was completed on May 28, 1989.

' 2.2 Emeroency Core Coolina System (ECCS) 2.2.1 Introduction On July 23, 1987, the BWROG submitted LTR NEDC-30936P (Ref. 3), "BWR Owner's Group Technical Specification Improvement Methodology (with Demonstration for BWR ECCS Actuation Instrumentation) Part 2," for staff review. This Topical Report utilizes fault tree methodology similar to that described in section 2.1.1. above to estimate the impact of the proposed changes on the average water injection function failure frequency utilizing only the Loss of Offsite Power (LOSP).

In a letter to the BWROG dated December 9, 1988, the staff forwarded its SER on the GE Topical Report NEDC-30936P. The staff's review of the Topical Report focused on the acceptability of extending the STIs from 31 days to 92 days, the A0T for surveillance from 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and the A0T for repair from I hour to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The staff approved the analytic methods and general conclusions of the LTR indicating that licensees must confirm the LTR applicability to the specific plant and address instrument drift.

2.2.2 Technical Specifications The licensee proposed the following changes to the ECCS Actuation TS which have generally been reviewed and approved in the staff's SER dated December 9, 1988 on LTR NEDC-30936P:

2.2.2.1 Allowed Out-of-Service Times A. Technical Specification Changes The A0Ts of TS Table 3.3.3-1, " Emergency Core Cooling System Actuation Instrumentation," Actions 30, 33, 36, 37, and 40 are being increased from I hour to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; and Action 32 is being identified as 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Action 35 is changed from 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The surveillance A0T of footnote (a) to TS Table 3.3.3-1 is being increased from 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Actions 30, 32, 36, 37, and 40 of TS Table 3.3.3-1 are revised to resolve " loss-of-function" issues.

Action 35 of TS Table 3.3.3-1 increases the time to restore an inoperable channel to OPERABLE status from 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months i

Footnotes (e) and (f) of TS Table 3.3.3-1 are deleted and replaced by new footnote (e)

The licensee's analysis indicates that NEDC-30936-A and the staff's associated SER dated December 9, 1988, directly correspond to the CPS solid-state ECCS actuation instrumentation design; therefore, the LTR and staff's SER is directly applicable to CPS.

Based on the above, the staff finds the licensee's proposed TS change acceptable.

t

l -

2.2.2.2 Surveillance Test Intervals A. Technical Specification Changes The CHANNEL FUNCTIONAL TEST interval specified in TS Table 4.3.3.1-1,

" Emergency Core Cooling System Actuation Instrumentation Surveillance Requirements,"

is increased from monthly (M) to quarterly (Q) for the following Trip Functions:

A.I.a -

Division I Trip System, RHR-A (LPCI Mode) and LPCS System, Reactor Vessel Water Level - Low Low Low, Level 1, A.I.b -

Drywell Pressure - High, A.I.c -

Reactor Vessel Pressure - Low (LPCI and LPCS Injection Valve Permissive),

A.I.d -

LPCI Pump A Start Time Delay Logic Card, A.I.e -

LPCS Pump Discharge Flow - Low, A.I.f -

LPCI Pump (A) Discharge Flow -- Low, A.2.a -

Automatic Depressurization System Trip System "1,"

ADS Logic "A" and "E," Reactor Vessel Water Level - Low Low Low, level 1, A.2.b -

Drywell Pressure - High, i

A.2.c -

ADS Timer, A.2.d -

Reactor Vessel Water Level - Low, Level 3, A.2.e -

LPCS Pumn Discharge Pressure - High, A.2. f -

LPCI Pump Discharge Pressure - High, A.2.g -

ADS Drywell Pressure Bypass Timer, A.2.h -

Manual Inhibit ADS Switch, B.I.a -

Division II Trip System, RHR B and C (LPCI Mode),

Reactor Vessel Water Level - Low Low Low, Lev-l 1, B.I.b -

Drywell Pressure - High, B.I.c -

Reactor Vessel Pressure - Low (LPCI Injection Valve Permissive),

B.1.d -

LPCI Pump B Start Time Delay Logic Card, B.I.e -

LPCI Pump (B) Discharge Flow - Low, B.I.f -

LPCI Pump (C) Discharge Flow - Low, B.2.a -

Automatic Depressurization System Trip System "2,"

ADS l

Logic "B" and *F," Reactor Vessel Water Level - Low i

Low Low, Level 1, B.2.b -

Drywell Pressure - High, i

B.2.c -

ADS Timer, t

B.2.d -

Reactor Vessel Water Level - Low, level 3, B.2.e -

LPCI Pump (B and C) Discharge Pressure - High, B.2.f -

ADS Drywell Pressure Bypass Timer, B.2.g -

Manual Inhibit ADS Switch, C.1.a -

Division III Trip System, HPCS System, Reactor Vessel l

Water Level - Low Low, Level 2, C.I.b -

Drywell Pressure - High, C.I.c -

Reactor Vessel Water Level - High, Level 8, i

C.I.d -

RCIC Storage Tank Level - Low, 1

?

l t

i C.I.e -

Suppression Pool Water Level - High, l

C. l. f.-

HPCS Pump. Discharge Pressure - High, and.

C.I.g -

HPCS System Flow Rate - Low i

In addition, the analog trip module calibration interval specified by i

footnote (a) to TS Table 4.3.3.1-1 is increased from 31 days to 92 days.

B. Instrument Drift i

The licensee indicated in its safety analysis that the current ECCS actuation instrumentation setpoint calculations include the effects of-instrument drift over surveillance intervals of 18 months for-all 4

instrument loops with the exception of the analog trip modules. To address the drift issue with the ECCS actuation analog trip modules, IP indicated that it reviewed the results.of_ monthly-calibration checks i

perfomed over a 1-year ~ period. As a result of this review, IP-concluded that. the quarterly drift is within present tolerances to allow -

an increase in STI for the ECCS actuation analog trip' modules.from 31 days to 92 days. Records showing the actual setpoint calculations and i

supporting data should be retained on-site for possible-future staff audit.

~'

The licensee's analysis of actual surveillance data appears to bound the.

potential for both linear and non-linear drift problems.

In addition, the licensee indicated that plant procedures require operators to

-l perform a CHANNEL CHECK at least once per every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for instruments 1

with redundant channels. Many instruments, _ including the RPS-analog trip modules, have redundant channels; therefore, should ' drift begin to-i occur, the licensee has the capability to identify the problem early and 1

take appropriate corrective action.

i Based on the licensee's analysis of instrument' drift, the verification of fault tree modeling data and direct LTR applicability, and-the-acceptability _

of the proposed changes, the staff concludes that-sufficient analysis has been 3

performed to support the proposed RPS STI increases' indicated above.

t 2.2.2.3 Editorial Technical Specification Changes An editorial-change deleting footnote "*" associated with Surveillance -

Requirement 4.3.3.2 is. acceptable because it was applicable until the-first.

refueling; outage which was completed on May 28, 1989.

l 2.3 Control' Rod-Block System i

2.3.1' Introduction On June 23, 1986,- the'BWROG submitted LTR NEDC-30851P, Supplement 1, l

" Technical Specification Improvement Analysis for BWR Control' Rod Block t i

Instrumentation,"'(Ref.~4).for NRC review. The LTR provides=a basis for extending the STI for control rod block function (CRBF) instrumentation. The 1

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._:~

~

=-.

?

^ i CRBF shares common instrumentation with the RPS. To gain the full safety benefit of the proposed RPS instrumentation TS changes, a corresponding STI extension for the CRBF is necessary.

Unlike the analyses discussed in sections 2.1 and 2.2 above, no specific i

fault trees were developed for the CRBF instrumentation.

Instead, the.

impact on the average control rod block failure rate was estimated based j

upon the results in NEDC-30851P-A (Ref. 1) (See section 2.1.) NEDC-30851P concluded that the benefits associated with the proposed changes to the RPS l

instrumentation TS offset any potential negative impact of extending the i

control rod block instrumentation surveillance intervals._

i In a letter dated September 22, 1988, the staff forwarded its SER on NEDC-30851P, Supplement 1.

The staff approved the analytic methods and general conclusions of-the LTR indicating that licensees must confirm the LTR applicability to the specific plant and address instrument drift.

2.3.2 Technical Specifications The licensee proposed the following changes to the CRBF instrumentation TS.

which have been reviewed and generically approved in the staff's SER dated l

September 22, 1988, on LTR NEDC-30851P, Supplement 1:

j 2.3.2.1 Allowed Out-of-Service Times j

The surveillance A0T of footnote (e) to TS Table 3.3.6-1, " Control Rod Block Instrumentation," is being increased from 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l

- Action 64 to TS Table 3.3.6-1 is being added to permit " loss-of-l function" for the scram discharge. level " Water Level-High," and the -

reactor coolant system recirculation flow " Upscale."

l The licensee's analysis indicates that NEDC-30851P-A, NEDC-30851P-A, f

Supplement 1, and the staff's associated SER dated ~ September 22,.1988 directly correspond to the CPS solid-state RPS CRBF instrumentation design, therefore, the LTR and staff's -SER is directly applicable to CPS.

Based on l

~ the above, the staff finds the-licensee's proposed.TS change acceptable.

f 2.3.2.2 Surveillance Test Intervals A. Technical Specification Changes

The CHANNEL FUNCTIONAL TEST interval specified in TS: Table 4.3.6-1, i

" Control Rod Block Instrumentation Surveillance Requirements," is i

increased from monthly (M) to: quarterly (Q) for the following Trip.

i Functions..

~

1.a Rod Pattern Control System, Low Power Setpoint, 1.b Rod Pattern Control System, RWL High. Power Setpoint, t

-t i

a

I t

2.a APRM Flow Biased Neutron Flux - Upscale, 2.b APRM Inoperative, t

2.c APRM Downscale, 2.d APRM Neutron Flux - Upscale, Startup, 5.a Scram Discharge Volume, Water Level - High, and 6.a Reactor Coolant System Recirculation Flow, Upscale In addition, the analog trip module calibration Nterval specified by footnote (f) to TS Table 4.3.6-1 is increased from 31 days to 92 days and note (d) is deleted to resolve conflicts with earlier submittals.

l B. Instrument Drift The licensee indicated in its safety analysis that the current RBCF instrumentation setpoint calculations include the effects of instrument drift over surveillance intervals of 18 months for all instrument loops with the exception of the analog trip modules. To address the drift issue with the RBCF instrumentation analog trip modules, IP indicated that it reviewed the results of monthly calibration checks performed over a 1-year period. As a result of this review, IP concluded that the quarterly drift is within present tolerances to allow an increase in STI for the RBCF instrumentation analog trip modules from 31 days to 92 days.

Records showing the actual setpoint calculations and supporting data should be retained on-site for possible future staff audit.

The licensee's analysis of actual surveillance data appears to bound the l

potential 1

for both linear and non-linear drift problems.

In addition, the licensee indicated that plant procedures require operators to perform a CHANNEL CHECK at least once per every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for instruments with redundant channels.

Many instruments, including the RBCF instrumentation analog trip modules, have redundant channels, therefore, should drift begin to occur, the licensee has the capability to identify the problem early and take appropriate corrective action.

Based on the licensee's analysis of instrument drift, the verification of l

fault tree modeling data and direct LTR applicability, and the acceptability of the proposed changes, the staff concludes that sufficient analysis has been performed to support the proposed CRBF instrumentation STI increases indicated above.

2.4 Containment and Reactor Vessel Isolation Control System (CRVIS) 2.4.1 Introduction On August 29, 1986, the BWROG submitted LTR NEDC-30851P, Supplement 2,

" Technical Specification Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation," for staff review.

Additionally, on June 27, 1989, the BWROG submitted LTR NEDC-31677P,

. ~.

-.. -.. ~.

. " Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation," for staff review. NEDC-30851P, Supplement 2 provides the i

basis for extending CRVIS instrumentation A0Ts and STIs for isolation instrumentation common to both ECCS and RPS. NEDC-31677P provides the basis for extending CRVIS instrumentation A0Ts and STIs for isolation instrumentation not common to RPS or ECCS.

In a letter to the BWROG dated January 6,1989, the staff forwarded its SER regarding the review of NEDC-30851P, Supplement 2.

The staff's evaluation concluded that NEDC-30851P, Supplement 2 provided an acceptable basis for extending STIs and A0Ts associated with isolation instrumentation common to RPS and ECCS.

In a letter to the BWROG dated June 18, 1990, the staff forwarded its SER regarding the review of NEDC-31677P. The staff concluded in its evaluation that NEDC-31677P provided an acceptable basis for extending STIs and A0Ts for isolation instrumentation not common to the RPS or ECCS.

2.4.2 Technical Specifications The licensee proposed the following changes to the CRVIS instrumentation TS (TS 3/4.3.2) which have been reviewed and generically approved in the staff's SERs dated January 6,1989 on LTR NEDC-30851P, Supplement 2, and June 18, 1990 on LTR NEDC-31677P:

2.4.2.1 Allowed Out-of-Service Times A. Technical Specification Changes The repair A0T of Action b.2 is increased from I hour to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The repair A0T of Action c.] is increased from I hour to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The repair A0T of footnote "**" associated with this Action is also increased from 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The surveillance A0Ts identified in footnote "*" to the Limiting Condition for Operation (LCO) and footnote (a) and-(k) to TS Table 3.3.2-1, "CRVIS Instrumentation," are increased from 2 hor ; to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The licensee's analysis indicates that NEDC-30851P, Supplement 2, and NEDC-31677P in addition to the staff's associated SERs dated January 6, 1989, and June 18, 1990,- directly correspond to the CPS solid-state RPS CRVIS instrumentation design; therefore the LTRs and staff's SERs are directly e

applicable to CPS.

Based on this affirmation, the staff finds the'11censee's proposed TS change acceptable.

2.4.2.2 Surveillance Test Intervals A. Technical Specifications

. The CHANNEL FUNCTIONAL TEST interval specified in TS Table 4.3.2.1-1,

CRVIS Instrumentation Surveillance Requirements,"

is increased from monthly (H) to quarterly (Q) for the following Trip Functions:

1.a Primary and Secondary Containment Isolation, Reactor Vessel Water Level - Low Low, level 2, 1.b Reactor Vessel Water Level - Low Low, Level 2 (ECCS Div. I and II),

1.c Reactor Vessel Water Level - Low Low, Level 2 (ECCS Div.

III),

1.d Drywell Pressure - High, 1.e Drywell Pressure - High (ECCS Div. I and II),

l.f Drywell Pressure - High (ECCS Div. III),

l.g Containment Building Fuel Transfer Pool Ventilation Plenum Radiation - High, l.h Containment Building Exhaust Ventilation Radiation - High, 1.1 Containment Building Continuous Containment Purge (CCP)

Exhaust Radiation - High, 1.j Reactor Vessel Water Level - Low Low Low, Level 1, 1.k Containment Pressure - High, 1.1 Main Steam Line Radiation - High, l.m Fuel Building Exhaust Radiation - High, 2.a Main Steam Line Isolation, Reactor Vessel Water Level - Low Low Low, Level 1, 2.b Main Steam Line Radiation - High, 2.c Main Steam Line Pressure - Low, 2.d Main Steam Line Flow - High, 2.e Condenser Vacuum - Low, 2.f Main Steam Line Tunnel Temp. - High, 2.g Main Steam Line Tunnel Delta Temp. - High, 2.h Main Steam Line Turbine Bldg. Temp. -High, 3.a Reactor Water t'leanup System Isolation, Delta Flow - High, 3.b Delta Flow Timer, 3.c.1 -

Equipment. Area Temp. - High, Pump Rooms A, B,-C, 3.c.2 -

Equipment Area Temp. - High, Heat Exchanger Rooms - East,

West, 3.d.1 -

Equipment Area Delta Temp. - High, Pump Rooms A, B, C, 3.d.2 -

Equipment Area Delta Temp. - High, Heat Exchanger Rooms -

East, West, 3.e Reactor Vessel Water Level - Low Low, Level 2, 3.f Main Steam Line Tunnel Ambient Temp. - High, 3.g Main Steam Line Tunnel Delta Temp. - High, 3.h SLCS Initiation, Reactor Core Isolation Cooling System Isolation, RCIC Steam 4.a Line Flow - High, 4.b RCIC Steam Line Flow - High Timer, RCIC Steam Supply Pressure - Low, 4.c 4.d RCIC Turbine Exhaust Diaphragm Pressure - High, 4.e RCIC Equipment Room Ambient Temperature - High, 4.f RCIC Equipment Room Delta Temperature - High,

t 4.g Main Steam Line Tunnel Ambient Temp. - High, 4.h Main Steam Line Tunnel Delta Temp. - High, 4.1 Main Steam Line Tunnel Temp. Timer, 4.j Drywell Pressure, i

4.1 RHR/RCIC Steam Line Flow - High, 4.m RHR Heat Exchanger A, B Ambient Temperature - High, 4.n RHR Heat Exchanger A, B Delta Temp. - High, 5.a kHR System Isolation, RHR Heat Exchanger Rooms A, B Ambient Temp. - High, 5.b RHR Heat Exchanger Rooms A, 8 Delta Temp. - High, 5.c Reactor Vessel Water Level - Low, Level 3, 5.d Reactor Vessel Water Level - Low Low Low, tevel 1, 5.e Reactor Vessel (RHR Cut-in Permissive) Pressure - High,

5. f.1 -

Drywell Pressure - High, RHR Test Line, and 5.f.2 -

Drywell Pressure - High, Fuel Pool Cooling The staggered test interval specified by footnote (a) to TS Table 4.3.2.1-1 is increased from 31 days to 92 days.

The analog trip module calibration interval specified by footnote (b) to TS Table 4.3.2.1-1 is increased from 31 days to 92 days.

B. Instrument Drift The licensee indicated in its safety analysis that the current CRVIS instrumentation setpoint calculations include the effects _ of instrument drift over surveillance intervals of 18 months for all instrument loops with the exception of the analog trip modules. To address the drift issue with the CRVIS instrumentation analog trip modules, IP indicated that it reviewed the results of monthly calibration checks performed over a 1-year period. As a result of this review, IP concluded that the quarterly drift is within present tolerances to allow an increase in STI for the CRVIS instrumentation analog trip modules from 31 days to 92 t

days.

Records showing the actual setpoint calculations and supporting data should be retained on-site for possible future staff audit.

The licensee's analysis of actual surveillance data appears to bound the potential for both linear and non-linear drift problems.

In addition, the licensee indicated that plant procedures require operators to perform a CHANNEL CHECK at least once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for instruments with redundant channels. Many instruments, including some of the CRVIS-instrumentation analog trip modules, have redundant channels, therefore, should drift begin to occur, the licensee has the capability to identify i

the problem early and take appropriate corrective action.

2.1.2.3 Editorial Technical Specification Changes An editorial change deleting footnote "**" associated with Surveillance-Requirement 4.3.2.2 is acceptable because it was applicable until the first refueling outage which was completed on May 28, 1989.

i o 2.5 Other Technical Specification Instrumentation On February 19, 1991 the BWROG submitted LTR GENE-770-06-1, " Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," for staff review. The staff forwarded its SER in a letter to the BWROG dated September 13, 1991 By letter dated July 21, 1992, the staff forwarded its SER regarding GENE-770-06-2 to the BWROG. The staff's SER concluded that GENE-770-06-2 provided sufficient basis to generically extend the A0Ts and STIs for selected i

instrumentation described below.

2.5.1 Technical Specifications The licensee proposed the changes to the following instrumentation TS which have been reviewed and generically approved in the staff's SERs dated September 13, 1991, and July 21, 1992, for LTR GENE-770-06-2 and LTR GENE-770-06-1, respectively:

l.

TS 3/4.3.4.1, "ATWS Recirculation Pump Trip System Instrumentation" a.

The surveillance A0T of footnote (a) to TS Table 3.3.4.1-1, "ATWS Recirculation Pump Trip System Instrumentation," is increased from 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

The CHANNEL FUNCTIONAL TEST interval specified in TS Table 4.3.4.1-1, "ATWS Recirculation Pump Trip Actuation Instrumentation Surveillanca Requirements," is increased from monthly (M) to quarterly (Q) for the following Trip Functions:

1 - Reactor Vessel Water Level - Low Low, level 2, and 2 - Reactor Vessel Pressure - High c.

The trip unit calibration interval specified by footnote "*"

to TS Table 4.3.4.1-1 is being increased from 31 days to 92 i

days.

2.

TS 3/4.3.4.2, "End-of-Cycle Recirculation Pump Trip System Instrumentation" a.

The repair A0T of Action c is increased from I hour to 6

hours, b.

The surveillance A0T of footnote (a) to TS Table 3.3.4.2-1, "End-of-Cycle Recirculation Pump Trip System Surveillance Requirements," is increased from 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l

i

_14_

c.

The CHANNEL FUNCTIONAL TEST interval specified in TS Table 4.3.4.2-1, "End-of-Cycle Recirculation Pump Trip System Surveillance Requirements," is increased from monthly (M) to quarterly (Q) for the following Trip Functions:

1 - Turbine Stop Valve - Closure, and 2 - Turbine Control Valve - Fast Closure d.

An editorial change deletes the " Total Number of Channels" and the " Channels to Trip" columns of TS Table 3.3.4.2-1 and revises the Minimum OPERABLE Channels per Trip Function requirements from "3" to "4."

Since this logic is arranged i

in a two-out-of-four.cheme and the Action Statements address channel inoperability based on four channels, this i

proposed change is being made to make Table 3.3.4.2-1 match the current Action Statements. This change does not reduce l

the number of channels required to be OPERABLE.

e e.

An editorial change deleting footnote "*" associated with Surveillance Requirement 4.3.4.2.2 is acceptable because it was applicable only until the completion of the first refueling outage which was complete on May 28, 1989.

[

3.

TS 3/4.3.5, " Reactor Core Isolation Cooling System Actuation Instrumentation" a.

The repair A0Ts of TS Table 3.3.5-1, " Reactor Core Isolation Cooling System Actuation Instrumentation," Actions 50.a and 52 are increased from 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; and Action 51 is identified as 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and Action 53 is increased from t

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

The surveillance A0T of footnote (a) to TS Table 3.3.5-1 is increased from 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.

The CHANNEL FUNCTIONAL TEST interval specified in TS Table l

4.3.5.1-1, " Reactor Core Cooling Actuation lastrumentation Surveillance Requirements," is increased from monthly (M) to quarterly (Q) for the following Functional Units:

1 - Reactor Vcssel Water Level - Low Low, Level 2, t

2 - Reactor Vessel Water Level - High, Level 8, 3 - RCIC Storage Tank Level - Low, and

't 4 - Suppression Pool Water Level - High d.

The analog trip module calibration interval specified by footnote (a) to TS Table 4.3.5.1-1 is increased from 31 days to 92 days.

l f

e 1

e.

Actions 50 and 52 of TS Table 3.3.5-1 have been revised to address " loss-of-function."

I f.

An editorial change deletes footnote "*" associated with Surveillance Requirement 4.3.5.2 since this footnote was only applicable until the first refueling outage which was completed on May 28, 1989.

4.

TS 3/4.3.A, " Control Rod Block Instrumentation" a.

Action 64 is added to TS Table 3.3.6-1, " Control Rod Block i

Instrumentation," to allow control rod block instrumentation channels to be inoperable for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months rather than I hour per existing Action 62. The reference to Action 62 on Table 3.3.6-1 is being replaced with a reference to Action 64 for the following Trip Functions:

5.a - Scram Discharge Volume, Water Level - High, and 6.a - Reactor Coolant System Recirculation Flow, Upscale.

5.

TS 3/4.3.7.1, " Radiation Monitoring Instrumentation" a.

The repair A0T of TS Table 3.3.7.1-1, " Radiation Monitoring Instrumentation," Action 70.a is increased from I hour to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

6.

TS 3/4.3.9, " Plant Systems Actuation Instrumentation" a.

Action Statement "b" has been deleted and replaced with new Action Statements 52 and 53 for Table 3.3.9-9, " Plant Systems Actuation Instrumentation." The proposed Action 52 is based on the number of channels affected instead of the number of loops affected.

Proposed Action 53 applies to those trip functions with only one channel per containment spray loop. A repair A0T extension to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is also applimble in those instances where taking out the channel dor cause a loss of trip function.

j Ado? lunally, Action 50 is revised to address the~ number of channels affected and to remove the option of placing the i

trip system in the tripped condition to satisfy the action requirement. Action 50 is also revised to be applicable to i

the Suppression Pool Water Level-Low-Low Trip function,-

Table 3.3.9-1 function 3.c.

An A0T extension to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months has been added for those situations where taking out the channel does not cause a loss of trip function.

Action Statement "c" is relocated to Table 3.3.9-1 as new Action 54.

e o Action Statement "d" is replaced with new Action Statement "b" which directs the action to be taken as required by Table 3.3.9-1.

b.

The surveillance A0T of footnote "*" to TS Table 3.3.9-1 is i

increased from 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The reference to " Trip System" is replaced with the term " Channel" to reflect the revised Action Statement terminology.

c.

The CHANNEL FUNCTIONAL TEST interval specified in TS Table 4.3.9.1-1, " Plant Systems Actuation Instrumentation Surveillance Requirements," is increased from monthly (M) to quarterly (Q) for the following Trip Functions:

1.a - Containment Spray System, Drywell Pressure - High, 1.b - Containment Pressure - High, 1.c - Reactor Vessel Water Level - Low Low Low, Level 1, l.d - Timers, 2.a - Feedwater System / Main Turbine Trip System, Reactor Vessel Water Level - High, Level 8, 3.a - Suppression Pool Makeup, Drywell Pressure - High, 3.b - Reactor Vessel Water Level - Low Low Low, Level 1, 3.c - Suppression Pool Water Level - Low Low, and 3.d - Suppression Pool Makeup Timer.

d.

The analog trip module calibration interval specified by footnote (a) to TS Table 4.3.9.1-1 is increased from 31 days to 92 days.

e.

The analog comparator unit calibration interval specified by footnote (b) to TS Table 4.3.9.1-1 is increased from 31 days to 92 days.

f.

An editorial change to correct the spelling of the word "comparator" in footnote (b) to TS Table 4.3.9.1-1.

g.

An editorial change to Action a to make it consistent with the wording utilized in similar Action Statements of other TSs.

h.

An editorial change to delete footnote "*" associated with-Surveillance Requirement 4.3.9.2 since this footnote was only applicable until the first refueling outage which was completed on May 28, 1989.

7.

TS 3/4.4.2.1, " Safety / Relief Valves" a.

The surveillance A0T of footnote "*" is increased from 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

-~

n i*

/

o,

b.

The CHANNEL FUNCTIONAL TEST interval of Surveillance Requirement 4.4.2.1.2.a is increased from 31 days to 92 days.

c.

An editorial change to delete footnote "**" associated with Surveillance Requirement 4.4.2.1.2.b since this footnote was only applicable until the first refueling outage which was

{

complete on May 28, 1989.

e 8.

TS 3/4.4.2.2, " Safety / Relief Valves low-Low Set function" a.

The surveillance A0T of footnote "*" is increased from 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

The CHANNEL FUNCTIONAL TEST interval of Surveillance Requirement 4.4.2.2.a is increased from 31 days to 92 days.

c.

An editorial change to delete footnote "+" associated with Surveillance Requirement 4.4.2.2.b since this footnote was only applicable until the first refueling outage which was i

complete on May 28, 1989.

2.5.2 Instrument Drift The licensee indicated in its safety analysis that the current instrumentation setpoint calculations include the effects of instrument drift over surveillance intervals of 18 months for all instrument loops with the exception of the analog trip modules. To address the drift issue with other instrumentation analog trip modules, IP indicated that l

)

it reviewed the results of monthly calibration checks performed over a 1-year period. As a result of this review, IP concluded that the quarterly drift is within present tolerances to allow an increase in STI for the instrumentation analog trip modules as indicated in the above changes.

Records showing the actual setpoint calculations and supporting data should be retained on-site for possible future staff audit.

i The licensee's analysis of actual surveillance data appears to bound the 2

potential for both linear and non-linear drift problems.

In addition, the licensee indicated that plant procedures require operators to perform a CHANNEL CHECK at least once per every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for instruments with redundant channels. Many instruments, including some of the.

instrumentation analog trip modules, have redundant channels; therefore, should drift begin to occur, the licensee has the capability to identify the problem early and take appropriate corrective action.

3.0 STATE CONSULTATION

In accordance with the Commission's regulations, the appropriate Illinois State official was notified of the proposed issuance of the amendment. The State official had no comments.

i

n

)?

1

/ 1

4.0 ENVIRONMENTAL CONSIDERATION

This amendment changes requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or changes surveillance requirements. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (56 FR 66921). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

5.0 CONCLUSION

The staff has concluded, based on the considerations discussed.above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: Anthony T. Gody, Jr., NRR i

Clifford Doutt, SICB Carl Schulten, OTSB Date: April 9, 1993 f

I a

ML20035F646

Contents

Text

1.0 INTRODUCTION

3.0 STATE CONSULTATION

4.0 ENVIRONMENTAL CONSIDERATION

5.0 CONCLUSION

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ML20035F646

Text

1.0 INTRODUCTION

3.0 STATE CONSULTATION

4.0 ENVIRONMENTAL CONSIDERATION

5.0 CONCLUSION

Navigation menu

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