JSP-609-91, Application for Amend to License NPF-62,changing Tech Specs to Incorporate reliability-based Improvements to Instrumentation Action Statements & Surveillance Test Intervals Per BWR Owners Group Topical Repts

Application for Amend to License NPF-62,changing Tech Specs to Incorporate reliability-based Improvements to Instrumentation Action Statements & Surveillance Test Intervals Per BWR Owners Group Topical Repts

ML20078A054
Person / Time
Site:	Clinton
Issue date:	09/20/1991
From:	Jamila Perry ILLINOIS POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20078A058	List: JSP-609-91, Application for Amend to License NPF-62,changing Tech Specs to Incorporate reliability-based Improvements to Instrumentation Action Statements & Surveillance Test Intervals Per BWR Owners Group Topical Repts U-601871, Proposed Tech Specs Re Reactor Protection Sys Instrumentation,Containment & Reactor Vessel Isolation Control Sys & ECCS Instrumentation Surveillance Requirements
References
JSP-0609-91, JSP-609-91, U-601871, NUDOCS 9109300199
Download: ML20078A054 (36)

Text

~

(' w b .

minois Power Company g ,-

Chnton Power Station P O. Dot 678

,,. Chnton. lL 61727 Tel ?17 935 6226 J. Stephen Perry Wee Pres &nt It. Lipid >I5i u-601871 L f>6/WW1 Eft L47-91 (09 -3 0 ) te 8E.100a JSP- 0609-91 September 20, 1991 10CFR50.90 Docket No. 50-461 Document Control Desk Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

-Clinton Power Station Proposed Amendment of Facility l Operatina License No. NPF-E2 l

De:r Sir:-

l PurWOLnt to 10CFR50.90, Illinois Power Company (IP) hereby applias for an.endment of Facility Operating License No. NPF-  !

62, 7.ppcndik A - Technical Specifications, for Clinton Power i Stat ton - (CPS) . This request consists of proposed changes to the CPA' Technical Specifications to incorporate reliability-based japrovements to instrumentation Action Statements and surveillance test intervals based on Topical Reports which  ;

have previously been submitted to the NRC by the Boiling Nater Reactor Owners' Group (BWROG). It should be noted that several of the proposed changes are based on one Topical Report (GENE-770-06-1) which has not yet been approved _by=the.NRC. Howover, as described in Attenhment 2,

, these proposed changes _are bounded by the analyses provided in the Topical Reports which have already been approved by the NRC.- IP is requesting these additional changes at this time to provide a complete request with respect to these reliability-based improvements. If the portion of this request based on GENE-770-06-1 is not approved, some of the remaining proposed changes (which are based on NRC-approved

-Topical Reports) would.not be able to be implemented. This is because these instruments perform multiple functions which are addressed by separate Technical Specifications and hence, are addressed by separate Topical Reports.

For each-of the above-noted proposed Technical Specification changes, a description and the associated justification (including a Basis For No Significant Hazards Consideration) are provided in Attachment 2. Marked-up copies of pages from the. current CPS isonnical Specificatione are provided in Attachraent 3. In addition, an affidavit supporting the facts Fat forth in this letter and its attachments is provided in Attachment 1.

9109300199 910920 'D

{ ge r DR ADOCK0500gggi l

4 IP has reviewed the proposed changes against the criteria of 10CFR51.22 for categorical exclusion from environmental impact considerations. The proposed changes do not involve a significant hazards consideration, or significantly increase the amounts or change the types of effluents that may be released offsite, nor do they significantly increase individual or cumulative occupational radiation exposures.

Based on the foregoing, IP concludes the proposed changes meet the criteria given in 10CFR51.22(c)(9) for a categorical exclusion from the requirement for an Environmental Impact Statement.

Sincerely yours,

\U )

f :SO y

[J. S. PerryQ V-i e Presidbnt DAS/alh Attachme:.'.s cc: NRC Clinton Licensing Project Manager NRC Resident Inspector, V-690 NRC Region III, Regional Administrator Illinois Department of Nuclear Safety

Attachment 1 y to U-601671 STATE OF ILLINOIS COUNTY OF DEWITT-J. StepL n Perry, being first duly sworn, deposes and says:

That he is Vice President of Illinois Power Company; that the application for amendment of Facility Operating License NPF-62 has been prepared under his supervision and direction; that he knows the contents thereof; and that to the best of his knowledge and belief said e.pplication and l the facts contained therein are true and correct.

DATE: This Ro day of September 1991.

Signed: h[ \ \

J tephnPerr d me Subsc. -ibed and sworn to befor/e diteiitis'4F day of

, ;l' L i,

7/[ ,

)4 A,

- u

, Notary Public l

l l

g L

I l

l

, Attochment 2 to U 601871 LS88-042 Page 1 of 33 packnround During late 1983, the BWR Owners' Group (BWROG) for med a Technical Specification Improvement (TSI) Committee, of which Illinois Power (IP) is a member. This committee subsequently established a program for the development of reliability analyses to justify improvements to surveillance test intervals (STIs) and allowable out of-service times (A0Ts) for instrumentation specified in the BWR Standard Technical Specifications. The primary objective of this program was to minimize, for applicable instrumentation, unnecessary testing and excessively restrictive A0Ts that could potentially degrade overall plant safety and availability. Examples of some of the problems experienced with the current Technical Specification requirements are: inadvertent scrams or engineered safety feature actuations due to frequent testing; A0Ts which are not long enough to perform repairs on a reasonable basis; excessive actuation of equipment for testing contributing to wear-out; and unnecessary radiation exposure to personnel performing Technical Specification required testing. A reduction in the number of Technical Specification required surveillance tests will allow plant personnel to perform other activities to increase the overall safety of the plant.

Within the same time frame, the NRC Staff issued NUREG-1024, " Technical Specifications - Enhancing the Safety Impact," which recommended that surveillance test requirements and Technical Specification Action Statements be reviewed to assure that they have an adequate technical basis and do indeed minimize plant risk. Use of reliability analyses to support engineering judgement was recognized as a primary basis for  ;

improving the Technical Specification requirements. NUREG-1024 thus reinforced the BWROG's program objectives and implementation wethodology.

To this end, the BVROG submitted a series of Licensing Topical Reports addressing the Technical Specification instrumentation requirements for the Reactor Protection System (NEDC-30851P), Emergency Core Cooling -

Systems (NEDC-30936P), the Control Rod Block System (NEDC-30851P, Supplement 1), and the Containment and Reactor Vessel Isolation Control System (NEDC-30851P, Supplement 2 and NEDC-31677P). Each of these Licensing Topical Reports has been reviewed and approved by the NRC. In addition, the BWROC has submitted a Licensing Topical Report (GENE-770-06-1) which addresses Technical Specification requirements for other instruments which are rimilar to those addressed in the Licensing Topical Reports previously reviewed and opproved by the NRC. However, GENE-770 06 1 has not yet been approved by the NRC.

As a member of the BWROG Technical Specifications Committee, IP is requesting that the results of the BVROG Licensing Topical Reports on Technical Specification improvements be applied to Clinton Power Station (CPS). For convenience in reviewing this request, this submittal (i.e .,

this attachment) has been divided into five separate parts addressing the functional areas and associated BWROG Licensing Topical Report (s).

Each part contains its own description of proposed changes, justification, and Basis for No Significant Hazards Consideration.

Included in Attachment 3 are marked-up copies of pages from the current CPS Technical Specifications indicating the combined effect of the changes requested in each part of ibis attachment.

, Attachment 2 to U-601871 LS-88 042 Page 2 of 33 Fort I - Reactor Protection System (RPS)

Description of Proposed C,bances In accordance with 10CFR50.90, the following changes to Technical Specification 3/4.3.1, " Reactor Protection System Instrumentation," are proposed:

1. The repair allowable out-of service time (AOT) of Action a.2 is being increased from one hour to six hours.

2. The surveillance A0T of footnote "*" is being increased from two hours to six hours.

3. The CHANNEL FUNCTIONAL TEST interval specified on Technical Specification Table 4.3.1.1-1, " Reactor Protection System Instrumentation Surveillance Requirements," is 'oeing increased from weekly (W) or monthly (M), as applicable, to quarterly (Q) for the following Functional Units;

a. item 2.b, Average Power Range Monitor (APRM) Flow-Biased Simulated Thermal Power - High,

b. item 2.c. APRM Neutron Flux - High,

c. item 2.d, APRM Inoperative,

d. item 3, Reactor Vessel Steam Dome Pressure - High,

e. Item 4, Reactor Vessel Water Level - Low, Level 3,

- f. item 5, Reactor Vessel Water Level - High, Level 8, g, item 6, Main Steam Line Isolation Valve - Closure,

h. item 7, Main Steam Line Radiation - High,

1. item 8, Drywell Pressure - High, J. item 9.a. Scram Discharge Volume Water Level - High, Level Transmitter,

k. item 10, Turbine Stop Valve - Closure,

1. item 11, Turbine Control Valve Fast Closure Valve Trip System 011 Pressure - Low, and

m. item 13, Manual Scram.

4. The analog trip module calibration-interval specified by footnote (g) to Technical Specificati s n Table 4.3.1.1-1 is being increased from 31 days to 92 days.

S. An editorial change is bel.ng proposed to delete footnote "**"

associated with surveillance Requirement 4.3.1.2 since this footnote was only applicable until the first refueling outage.

Justification for Proposed Channes on May 31, 1985 the BWROG submitted Licensing Topical Report NEDC-30851P, " Technical Specification Improvement Analyses for BWR Reactor Protection System," for NRC review. (This report provides justification for the proposed changes identified as 1 through 4 above. ) The analyses documented in NEDC-30851P utilized fault tree modeling (based upon the CPS design) to estimate the impact of the proposed changes on the average Reactor Protection System (RPS) failure frequency.

i . . . . ps

.. Attachment 2

. g. ,

to U-601871.

LS88-042 Page 3 of'33 LThe average RPS failure frequency'is a function of the frequency of scram-demands and the probability that the RPS is-unavailable when

-demanded. The initiating events which require successful operation of

.the' RPS for ensuring safe reactor shutdown were identified and their

-annual occurrence frequencies were estimated, The initiating events were divided into three groups based on the number of diverse seraors that initiate the scram for that event.

For each initiating event, a top-level failure event was identiffed using the success criteria described below. For each top failure event, a fault tree was developed which modeled all of the components needed-for generation and processing of the RPS signals including the sensors, analog trip modules, logic cards, load drivers and scram solenoids. The common cause failure of these components was also modeled. A fault tree analysis was then performed using the WAM series computer code, WAMCUT, to obtain the major failure cut sets that contribute to the top failure event probability. The failure cut sets 'obtained were then analyzed using the FRANTIC III computer code to determine the average RPS system unavailability upon demand.

The average RPS unavailability was calculated for each initiating event group based on inputs which included component failure rates (time'and

-demand related), common cause failure rates, human error rates, testing intervals, and test and repair times.- Sensitivity studies were conducted by changing the input parameters by f ac* ors of 2, 5 and 10 (and 30 where appropriate) to determine the resultant impact on the

-average RPS unavattability and the total RPS failure frequency. The surveillance test intervals (STIs) and A01s were then varied to determine the resulting effect on the average RPS failure frequency.

The scram success critoria used for this analysis is defined below for two specific failure modes:

a. Failure Mode A: One or more RPS electrical control rod-groups fail to insert into the core. The success criteria for this. failure mode was that two of the total four rod groups must fully insert.

b. Failure Mode B: One or more control rods in a random pattern fail to insert. The success criteria for this failure mode was that, if the control rods are inserted in a random manner, 69% of all the rods must fully insert to achieve success.

-The. acceptance guideline used by the BWROG for the proposed changes is based on a net change in risk. The net change in risk is the difference between the increase-in risk that would result from the-proposed changes and the decrease in risk that would result from the reduced likelihood oi Jnadvertent scrams. If the net: change in risk is. determined to be insignificant, tius BWROG considered the proposed changes to be acceptabic.

h 3 Attachment ?'

to U-601871 L

LS-88 042 L

• Page 4 of 33

[ :The BWROG concluded that the overall effect of the proposed RPS.

Technical Specification changes provides a net increase in safety and improves plant operation. Tha improvement is achieved by reducing the j potential for: (a) unnecessary plant scrams (reduced challenges to t plant shutdown systems and improved plant availability); (b) excessive o test cycles on equipment _(reduced wear out potential); and (c) diversion

[

D of plant personnel and resources on unnecessary testing (potential safety and operational improvement). The BWROG report concluded that

erage RPS failure frequency increases from

( forCPSthecalculatedag/yearanda0.1%increaseinplantcapacity 2.0x10-6 / year to 2.3x10" l-factor can be achieved with incorporation of the proposed RPS Technical Specification changes.

[ By letter from Ashok C. Thadani (NRC) to Robert F. Janacek (BWRW .tated

January 24, 1988, the NRC provided their Safety Esaluation Repop ( f

-NEDC-30851P. The NRC concluded in their Safety Evaluation Report that l NEDC-30851P applies directly to CPS and that the proposed changes would

have a negligible impact on plant risk. On this basis, the NRC i'

determined that these proposed changes are acceptable. However, the Staff identified that NEDC-30851P does not confirm that the allowable

[ calibration period for instrumentation used in the RPS (for example, the l solid-state analog trip units) can be extended from monthly to quarterly

[ without creating excessive drift. Therefore, the Safety Evaluation p Report states that the licensee must demonstrate, by use of current j drift information provided by the equipment. vendor er by use of plant- ,

I: specific data, that a change of the_fu..ctional test interval from monthly to quarterly can be supported.

I i With respect to the Staff's concern about instrument drift, the instrument setpoint calculations for the RPS instruments at CPS include  !

the effects of instrument drift over 18 months for all instrument loop i components except for the analog trip' modules. To address drift of the i j

analog trip modules, IP reviewed the results of monthly calabration checks performed over a one-year period on the-affected RPS analog trip j modules. _ Review of these calibration checks showed that the quarterly (

' drift is within the present calibration tolerances. As a result, IP has

-concluded that 1cngthenin6 the CHANNEL FUNCTIONAL TEST interval snd '

analog trip module calibration interval, as applicable, for the RPS '

instruments from weekly or monthly to quarterly will nor result in excessive instrument drift relative to the current, established setpoints. In addition, a CHANNEL CHECK is required at least once per .

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for- those instruments with redundant channels. These routine CHANNEL CHECKS will help to identify excessive drif t of the RPS j instrumentation.

1 Vith respect to NRC approval of plant-specific changes to the RPS Technical Specifications based upon NEDC-30851P, IP understands that-the- '

NRC .has expressed concern that the changes proposed in NEDC-30851P would

. allow continued plant operation for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with a combination of '

failures which could prevent a reactor scram as assumed for a particular l

_ plant-transient. -This c;ould occur for a relay-type plant (with one-out- (

of-two-twice logic) if, for example, both channels of high reactor  ;

pressure and both channels of APRM neutron flux-high were inoperable in i one trip system. These are the two RPS scram functions that are assumed  ;

to mitigate-the pressure regulator failure - increasing transient (see r I

, Attachment 2 to U 601871 LS88-042 Page 1 of 33 Teble F-1 ef NEDC-30951P). The prop:ced recolution to thic ic:ue is tc require the affectcd trip cyctea to be placed in the tripped condition within one hour (rather than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) when such a loss-of-function condition exists.

Because the solid-state RPS design of CPS is arranged in a two-out-of-four type logic scheme, this issue is not applicable to CPS. The proposed changes do not allow continued operation when any parameter is unable to provide a reactor scram. Therefore, rerolution of the " loss-of-function" issue for NEDC-30851P does not impact approval of this request for CPS.

Basis For No Sinnificant Hazards Consideration In accordance with 10CFR50.92, a proposed change to the operating license (Technical Specifications) involves no significant hazards considerations if operation of the facility in accordance with the proposed change would not: (1) involve a significant increase in the probability or consequences of any accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety. The proposed RPS Technical Specification changes are evaluated against each of these criteria below.

(1) These proposed changes do not involve a change to the plant design or operation, only to the allowable out-of-service times (A0Ts) and fiequency at which testing of the RPS instrumentation is perfotaed. As a result, these proposed changes cannot increase the probability of any design basis accident previously evaluated.

As identified in NEDC-30851P, these propos d changes increa e the averageRPSfailure{/

This increase (3x10~ year)isconsideredtoberequency insignificant.

from 2.0x10 As identified in the NRC Staff's Safety Evaluation Report of NEDC-30851P, this increase in average RPS failure frequency would contribute to a very small increase in core-melt freauency. The small increase in average RPS failure frequency is otract by safety benefits such as a reduction in the number of inadvertent test-induced scrams, a reduction in wear due to excessive equipment test cycling, and better optimization of plant personnel resources. Hence, the net change in risk resulting from these proposed changes would be insignificant. Therefore, these proposed changes do not result in a significant increase in the probability or the consequences of any accident previously evaluated.

(2) These proposed changes do not result in any change to the plant design or operation, only to the A0T and frequency at which testing of the RPS instrumentation is performed. Since failure of the RPS instrumentation itself cannot create an accident, these proposed changes can at mest af fect on.y accidents which have been previously evaluated. Therefore, these proposed changes cannot create the possibility of a new or different kind of accident from any accident previously evaluated.

, Attachment 2 I

co U-601871 13-88-042 Page 6 of 33 (3) As identified above, three prepened ebenges increcs the everage RPS failure frequency from 2.0x10'0/yearto2.3x10'g/ year. The NRC Staff's Safety Evaluation Report of NEDC-30851P cor.cluded that this small average RPS failure frequency increase would contribute to a very small increase in core-melt frequency. This small increase in average RPS failure frequency would be offset by safety benefits such as a reduction in the number of inedvertent test-induced scrams, a reduction in wear due to excessive equipment test cycling, and better optimization of plant personnel resources, Hence, the net change in risk resulting from these proposed changes would be insignificant. In addition, IP has confirmed that the proposed changes to the functional test intervals will not result in excessive instrument drift relative to the current, established setpoints. Therefore, these proposed changes do not result in a significant reduction in a margin of safety.

Based upon the foregoing, IP concludes that these proposed changes do not involve a significant hazards consideration.

Attachment 2

' to U 601871 LS88-042 Page 7 of 33 Part 11 crern-~ra cara rac14a- S'?*e? (ccci' pescription of Proposed Chanr.es In accordance with 10CER50.90, the following changes to Technical Specification 3/4.3.3, " Emergency Core Cooling System Actuation Instrumentation," are proposed:

1. The repair alloseble out-of-service times (A0Ts) of Technical Specification Table 3.3.3-1, " Emergency Core Cooling System Actuation Instrumentation," Actions 30, 33, 36, 37, and 40 are being increased from one hour to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; Action 35 is being increased from eight hours to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; and Action 32 is being identified as 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2. The surveillance AOT of footnote (a) to Technical Specification Tabic 3,3.3 1 is being increased from two hours to six hours.

3. The CRANNEL FUNCTIONAL '?EST interval specified on Technical Specification Table 4.3.3.1-1, " Emergency Core Cooling System Actuation Instrumentation Surveillance Requirements," is being increased from monthly (M) to quarterly (Q) for the following Trip Functions:

a. item A.l.a. Division i Trip System, RHR-A (LPCl Mode) and LPCS System, Reactor Vessel Water Level - Low Low Low, Level 1,

b. item A.l.b, Drywell Pressure - High,

c. item A.l.c, Reactor Vessel Pressure -

Low (LPCI and LPCS Injection Valve Permissive),

d. item A.l.d LPCI Pump A Start Time Delay Logic Card,

e. item A l.e, LPCS Pump Discharge Flow - Low,

f. Item A l.f, LPCI Pump (A) Discharge Flow - Low,

g. item A.2.a Automatic Depressurization System Trip System "1", ADS Logic "A" and "E", Reactur Vessel Water Level - Low Low Low, Level 1,

h. item A.2.b, Drywel! Pressure - High,

1. item A.2.c. ADS Timer, j Item A.2.d, Reactor Vessel Water Level - Low, Level 3,

k. Item A.2.e, LPCS Pump Discharge Pressure - High,

1. item A.2.f. LPCl Pump A Discherge Pressure - High,

m. item A.2.g, ADS Drywell Pressure Bypass Timer,

n. item A.2.h, Manual Inhibit ADS Switch,

o. ftem B.l.a. Division II Trip System, RHR B and C (LPCI Mode), Reactor Vessel Water Level Low Low Low, Level 1,

p. item B.l.b, Drywell Pressure - High,

q. item B.l.c, Reactor Vessel Pressure - Low (LPCI Inj ection Valve Permissive),

r. item B.l.d, LPCi Pump B Start Time Delay Logic Card,

s. Item B.l.e, LPCI Pump (B) Discharge Flow - Low,

t. item B.l.f, LPCI Pump (C) Discharge Flow - Low,

u. Item B.2.a, Automatic Depressurization System Trip System "2", ADS Logic "B" and "F", Reactor Vessel Water Level - Low Low Low, Level 1,

v. Item B.2.b, Drywell Pressure - High,

, Attachment 2 to U 601871 LS-88 042 l

. Page 8 of 33

w. item B.2.c, ADS Timer,

x. item B.2.d, Reactor Vessel Water Level Low, Level 3,

y. item B.2.e, LPCI Pump (B and C) Discharge Pressure - High, z, item B.2.f. ADS Drywell Pressure Bypese Timer, aa. item B.2.g, Manual Inhibit ADS Switch, bb. item C.l.a Division III Trip System. HPCS System, Reactor Vessel Water Level - Low Low, level 2, cc. item C.1.b, Drywell Pressure High, dd. item C.1.c, Reactor Vessel Vater Level - High, Level 8, ee. item C.l.d, RClc Storage Tank Level - Low, ff, item C.1.e, Suppression Pool Water Level - High, gg. item C.l.f, HPCS Pamp Discharge Pressure High, and hh. item C.1.g HPCS System Flow Rate - Low.

4. The analog trip module calibration interval specified by footnote (a) to Technical Specification Table 4.3.3.1-1 is being increased from 31 days to 92 days.

5. An editorial change is being proposed to delete footnote "*"

associated with Surveillance Requirement 4.3.3.2 since this footnote was only applicable until the first refueling outage.

Justification for Pronosed Channes On July 23, 1987 the BWROC submitted Licensing Topical Report NEDC-30936P, "BWR Owners' Group Technical Specification Improvement Methodology (with Demonstration for BWR ECCS Actuation Instrumentation)

Part 2," for NRC review. (This report provides justification for the proposed changes identified as 1 through 4 above.) Similar to the RPS report discussed in Part I of this submittal, the analyses documented in NEDC-30936P (Part 2) utilized fault tree modeling (based upon the CPS design) to estimate the impact of the proposed changes on the average water injection function failure frequency.

The calculation of average water injection failure frequency depends on two sets of parameters. The first set consists of initiating events which eventually call for water injection. The second set consists of the probability that the water injection function is unavailable given a demand for injection. Depend':6 on each initiating event, the number of components that are needed fot cuccessful completion of the water injection function varies. Ther; fore, the water injection unavailability for a given initiating event may differ from that of another initiating event.

A function fault tree was developed for each initiating event in order to quantify the water injection unavailability per demand. The function fault tree modeled the logical relationship of the faults that contribute to the water injection unavailability. The function fault tree was used to estimate the water injection unavailability based upon the current Technical Specification requirements and the effect of proposed changes. The results were considered acceptable by the BWROG if the proposed changes resulted in less than a 4% increase in the average water injection failure frequency or if the average water n failure frequency was calculated to be less than injectig/

1.0x10- year .

. Attachment 2

~' '

to U 601871 LS-88-042 Page 9 of 33 The only initiating evente studied in this analysia were less of offsite power (LOSP) initiating events, The LOSP event was chosen for conrideration because, based on prior Probabilistic Risk Assessment calculations, LOSP events contribute from 40% to 90% of the calculated core damage frequency for most BWRs. Also, the LOSP analysis is a more severe test of ECCS actuation instrumentation than other accident sequences such as turbine trip, loss of feedwater, and recirculation pump failure. Therefore, the effect of the proposed changes on water injection unavailability and failure frequency for the LOSP initiating event will dominate contributions from all initiating events.

By letter from Charles E. Rossi (NRC) to Donald N. Crace (BWROG) dated December 9, 1988, the NRC provided their Safety Evaluation Report of NEDC-30936P (Part 2). The NRC concluded in their Safety Evaluation Report that the methods and acceptance criteria provided in NEDC-30936P (Part 2) are acceptable for implementation on a plant-specific basis.

However, the NRC's Safety Evaluation Report states that in order for a licensee to use the generic analyses provided in NEDC-30936P (Part 2),

the licensee must confirm the applicability of the generic analyses to the plant and confirm that any increase in instrument drift due to the extended surveillance intervals is properly accounted for in the setpoint calculation methodology.

The CPS ECCS configuration was specifically modeled (identified as BWR 6 solid-state) in NEDC-30936P (Part 2). Therefore, the generic analyser are directly applicable to CPS. As identified on Tabic 3-1 of NEDC-30936P (Part 2), these proposed changes increase the ca culated average ection failure frequency for CPS from 7.69x10'J'/ year to wetering/

7.82x10' year. This represents an increase of 1.3x10 7/ year (1.7%),

which is well within the acceptance criteria of NEDC 30936P (Part 2) and the NRC's Safety Evaluation Report.

With respect to the NRC's concern about instrument drift, the instrument setpoint calculations for the ECCS actuation instrumentation at CPS include the effects of drift over 18 months for all instrument loop components except for the analog trip modules. To address drift of the analog trip modules, IP reviewed the results of monthly calibration checks performed over a one-year period on the affected ECCS actuation analog trip modules. Review of these calibration checks showed that the quarterly drift is within the present calibration tolerances. As a result, IP has concluded that lengthening the CHANNEL FUNCTIONAL TEST interval and analog trip module calibration interval, as applicable, for the ECCS actuation instruments from monthly to quarterly will not result in excessive drift relative to the current, established setpoints. In addition, a CHANNEL CHECK is required at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for those instruments with redundant channels. These routine CHANNEL CHECKS will help to identify excessive drift of the ECCS actuation instrumentation.

, Attachhent 2 l

to U 6(1871 LS 8P 042

. Pege 10 of 33 Basis For No Sinnificant Hazards Consideration i

in accordance with 10CFR50.92, a proposed change to the operating license (Technical Specifications) involves no significant hazards considerations if operation of the facility in accordance with the proposed change would not: (1) involve a significant increase in the probability or consequences of any accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety. The proposed ECCS actuation instrumentation Technical Specification changes are evaluated against each of these criteria below.

(1) These proposed changes do not involve a change to the plant design or operation, only to the allowable out-of service time (A0T) and frequency at which testing of the ECCS instrumentation is performed. Failure of the ECCS actuation instrumentation itself cannot create an accident. As a result, these proposed changes cannot increase the probability of any accident previously evaluated.

As identified in NEDC-30936P (Part 2), these proposed changes increase the alculated average water injection failure frequency from7.69x10'g/yearto7.82x10-6 / year. This represents an increase of 1.3x10'7/ year (1.7%), -

acceptancecriteria(4%or1.0x10'ghichiswellwithinthe

/ year) provided in NEDC-30936P (Part 2) and the NRC's Safety Evaluation Repor' This small increase in average water injection failuro frequency is offset by benefits such as a reduction in the number of inadvertent test-induced scrams and engineered safety feature actuations, a reduction in wear due to excessive test cycling, and bettcr optimization of plant personnel resources. Therefore, these proposed changes do not result in a significant increase in the consequences of any accident previously evaluated. -

(2) These proposed changes do not result in any change to the plant design or operation, only to the A0T and frequency at which testing of the ECCS instrumentation is performed. Since failure of the ECCS actuation instrumentation itself cannet create an accident, these proposed changes can at most affect only accidents which have been previously evaluated. Therefore, these proposed changes cannot create the possibility of a new or different kind of accident from any accident previously evaluated.

(3) As identified above, these proposed changes increase th d average water i jection failure irequency frcs calculatg/yearto7.82x10'g/

7.69x10' year. This increase is well eithin the acceptance criteria found acceptabic in the NRC Staf 's Safety Evaluation Report for NEDC-30936P (Part 2). Further, thi' small increase in average water injection failure frequency wou. i be offset by safety benefits such as a reduction in the numbe. of inadvertent test-induced scrams and engineered safety featni actuations, a reduction in wear due to excessive test cycling, , d better optimization of plant personnel resources. In addttion, IP has confirmed that the proposed changes to the functional test

. At tachmet.t 2 l ,

to U-601871 LS 88 042 Page 11 of 33 t

intervals will not result in excessive instrument drift relative to the current, established setpoints. Therefore, these proposed changes do not result in a significant reduction in a margin of safety.

Based on the foregoing, IP concludes that these proposed changes do not involve a significant hazards consideration.

9

-. Attachment 2-

, to U 601871 13 88 042 Pago 12 of 33 Part 111- Control-Rod B),,tLqh Rescription of Proposed Ch,amy,es -

In.accordance with-10CFR50.90, the followin6 changes to Technical

_ Specification 3/4.3.6, " Control Rod Block Instrumentation," are -

proposed *:

1. The surveillance allowable out-of-service time (AUT) of footnote (e) to: Technical Specification Table 3.3.6 1, " Control Rod Block .

Instrumentation," is beirg increased from two hours to six hours.

2. The CHANNEL FUNCTIONAL TEST interval specified. on Technical Specification Table 4.3.6-1, " Control Rod Block Instrumentation Surveillance Requirements," is being increased f rom monthly (M) to quartorly (Q) for the following Trip Functions:

a. Item 1.a. Rod Pattern Control System, Low Power Setpoint**,

b. iter 1,b, Rod Pattern Control System, RVL High Power Setpoint**,

c. item 2.a, APRM Flow Biased Neutron Flux - Upscale,

d. -item 2.b, APRM Inoperative,

.e. item ~2.c. APRM Downscale,

f. item 2.d, APRM: Neutron Flux - Upscale, Startup,

g. item 5.a, Scram Discharge Volume, Water-Level - High, and

h. item 6.a,. Reactor Coolant System Recirculation Flow, Upscale.

3. The analog-trip module calibration interval specified by footnote-(f) to Technical Specification Table 4.3.6-1 is being increased from 31 days to 92 days..

• -Additional changes to Technical Specification 3/4.3;6 are proposed -

in Part V of this submittal,

• • Currently, as indicated on the attached marked-up pageLfrom the CPS Technical Specifications-(page 3/4-3 68), two footnotes f(footnotes (d) and (e)] aro. attached to the CHANNEL FUNCTIONAL 1EST requirement for the RPCS 'LPSP and RVL HPSP on Table 4.3.6-1.

As'the required CHANNEL 1 FUNCTIONAL TEST frequency will be changed from_ monthly (M) to quarterly (Q) per.this reque,t,'it appears that footnote 1(d) will no longer be consistent since the purpose '

of the footnote was to modify the current' monthly (M) testE frequency requirement _by atter.hing the words, "at least once per-

-31 days while. operation continues within a given power range above the RPCS low power setpoint." Hewever, by IP letter dated. 0ctober 30, 1987:(reference U 601048), IP requested a change to' delete footnote-(d). With the approval of the october 30,-1987 ;equest.

no inconsistency should result from changing the CHANNEL FUNCTIONAL TEST frequency frce monthly (M) to quarterly (Q) per this request.

. Attachment 2 to U 601871 IS-88 042 Page 13 of 33 Justif.jcation for Proposed Chr.nnes on June 23, 1986 the BWROC submitted Licensing Topical Report NEDC-30851P, Supplement 1, " Technical Specification Improvement Analysis for BVR Control Rod Block Instrumentation," for NRC review. This report provides justification for each of the proposed changes identified above.

g Unlike the analyses discussed in Parts I and II of this submittal, no specific fault trees were dev91oped for the control rod block ins tr umenta t ion. Instead, the impact on the average control rod block failt.re rate was estimated based upon the results of the analyses presented in Part I of this submittal. This approach was taken because the heactor Protection System (RPS) and control rod block functions shara common instrument inputs. The BWROC report determined that the nge control rod block failure rate would increase less than ave 10' {/ year (0.06%)fromthecurrent failure rate of 0.16/ year (based on industry experience). NEDC-30851P, Supplement I states that the benefits associated with the proposed changes to the RPS instrumentation offset any potential negative impact of extending the control rod block instrumentation test intervals.

By letter-from Charles E. Rossi (NRC) to Donald N. Grace (bVROG) dated September 22, 1988, the NRC provided their Safety Evaluation Report of NEDC-30851P, Supplement 1. The NRC concluded in their Safety Evaluation Report that NEDC 30851P, Supplement 1 provides an acceptable basis for implementing the above proposed control rod block instrumentation changes. However, the NRC's Safety Evaluation Report states that in order for a licensee to use the generic analyses provided in NEDC 30851, Supplement 1, the licensee must confirm the applicability of the generic analyses to the plant and confirm that any increase in instrument drift due to the extended intervals is properly accounted for in the setpoint calculation methodology.

IP has confirmed that the control rod block inscrumentation configurat on (described in NEDC-30851P and Supplement 1 as the Rod i

Control and :nformation System) is identical to that at CPS. As a result, the ana?vses presented in NEDC-30851P, Supplement 1 are directly applicable to CPS.

With respect to the NRC's concern aoout instrument drift, the instrument setpoint calculations for the control rod block instrumentation at CPS include the effects of instrument drift over 18 months for all instrument loop components except for the analog trip modules. To address drift of the analog trip modules, IP reviewed the results of monthly calibration checks performed over a one-year period on the

=

affected control rod block analog trip modules. Review of these calibration checks showed that the quarterly drift is within the pecsent calibration tolerances. As a result. IP has concluded that lengthening the CllANNEL FUNCTIONAL TEST interval and analog trip module calibration interval, as applicable, for the control rod block instrumentation from monthly ta quarterly will not re s.u l t in excessive drift relative to the current, established setpoints. In addition, a CllANNEL CHECK is required at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for those instruments with redundant

. Attachment 2 to U-601871 LS88-042

- Page 14 of 33 channels. These routine CilANNEL CHECKS will help to identify excessive drift of the control rod block instrumentation.

Ea.s_i n For No Sirnificant Hazards Consideration In accordance with 10CFR50.92, a proposed change to the operating license (Technical Specifications) involves no significant hazards consideratiens if operation of the facility in accordance with the proposed change would not: (1) involve a significant increase in the probability or consequencer of any accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety. The proposed control rod block instrumentation Technical Specification changes are evaluated against each of these criteria below.

(1) These proposed changes do not involve a change to the plant design or operation, only to the allowable out of-service time (A0T) and frequency at which testing of the control rod block instrumentation is performed. Failure of the control rod block instrumentation itself cannot create an accident. As a result, these proposed changes cannot increase the probability of any accident previously evaluated.

As identified in NEDC-30851P, Supplement 1, these proposed changes increase the average control rod block failure frequency less than 0.06%. As provided in the NRC Staff's Safety Evaluation Report of NEDC-30851P, Supplement 1, this increase is very slight and is offset by the safety benefits associated with the proposed changes to the RPS instrumentation. As a result, the combined effect of the changes proposed for the RPS and control rod block instrumentation requirements should result in an overall improvement in plant safety. Therefore, these proposed changes do not result in c significant increase in the consequences of any -

accident previously evaluated.

(2) These proposed changes do not result in any change to the plant design or operation, only to the A0T and frequency at which testing of the control rod block instrumentation is performed.

~

Since failure of the control rod block instrumentation itself cannot create an accident, these proposed changes can at most affect only accidents which have been previously evaluated.

Therefore, these proposed changes cannot create the possibility of a new or different kind of accident from any accident previously evaluated.

(3) As identified above, these proposed changes increase the average control rod block failure frequency less than 0.06%. This increase is very slight and is offset by the safety benefits associated with the proposed changes to the RPS instrumentation.

As a result, the combined effect of the changes proposed for the RPS and control rod block festrumentation requirements should result in an overall improvement in plant safety. In addition, IP has confirmed that the proposed changes to the functional test intervals will not result in excessive instrument drift relative

. Attachment 2

, to U-601871 1.S88-042 Page 15 of 33 to the current, established setpoints. Therefore, these proposed changes do not result in a significant reduction in a margin of safety, Based on the foregoing, IP concludes that these proposed changes do not involve a r.ignificant hazards constderation.

, Attachment 2 to U-601871 LS 88 042

. Page 16 of 33 Part IV - Containment and Reactor Vessel Isolation Control System (CRVICS)

Description of Proposed Chances In accordance with 10CFR50.90, the following changes to Technical Specification 3/4.3.2, " Containment and Reactor Vessel Isolation Control System," are proposed:

1. The repair allowable out-of-service time (A0T) of Action b.2 is being increased from one hour to six hours.

2. The repair A0T of Action c.1 is being increased from one hour to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The repair A0T of footnote "**" associated with this Action is also being increased, from two hours to six hours.

3. The Surveillance A0Ts identified in footnote "*" to the Limiting Condition for Operation and footnotes (a) and (k) to Technical Specification Table 3.3.2 1, "CRVICS Instrumentation," are being increased from two hours to six hours,

4. The CHANNEL FUNCTIONAL TEST interval specified on Technical Specification Table 4.3.2.1-1, "CRVICS Instrumentation Surveillance Requirements," is being increased from monthly (M) to quarterly (Q) for the following Trip Functions:

a. item 1,a, Primary and Secondary Containment Isolation, Reactor Vessel Water Level - Low Low, Level 2,

b. item 1.b, Reactor Vessel Water Level - Low low, Level 2 (ECCS Div. 1 and II),

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

d. item 1.d, Drywell Pressure - High,

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

f. Item 1.1, Drywell Pressure - High (ECCS Div. III),

g. item 1.g, Containment Building Fuel Transfer Pool Ventilation Plenum Radiation - High,

h. item 1.h, Containment Building Exhaust Radiation - High,

i. item 1.1, Containment Building Continuous Containment Purge (CCP) Exhaust Radiation - High, J. item 1.j , Reactor Vessel Water Level - Low Low Low, Level 1,

k. item 1.k, Containment Pressure - High,

1. item 1.1, Main Steam Line Radiation - High,

m. Item 1.m. Fuel Building Exhaust Radiation - High, n, item 2.a, Main Steam Line Isolation, Reacte Vessel Water Level - Low Low Low, Level 1, o, item 2.b Main Steam Line Radiction - High,

p. item 2.c, Main Steam Line Pressure - Low,

q. item 2.d, Main Steam Line Flow - High,

r. item 2.e, Condenser Vacuum - Low,

s. Item 2.f Main Steam Line Tunnel Temp. - High,

t. Item 2.g, Main Steam Line Tunne; Delta Temp. - High,

u. iter. 2.h, Main Steam Line Turbine Bldg. Temp. - High,

. Attachment 2 to U 601871 LS-88 042 Page 17 of 33

v. item 3.a. Reactor Water Cleanup System Isolation, Delta Flow

- lii gh ,

v. item 3.b, Delta Flow Timer,

x. Item 3.c.1, Equipment Area Temp. lii gh , Pump Rooms - A , B , C ,

y. item 3.c.2, Equipment Area Temp. Hir,h, Heat Exchanger Rooms - East, West,

z. item 3.d.1, Equipment Area Delta Temp. liigh, Pump Rooms -

A,B,C, aa. item 3.d.2, Equipment Area Delta Temp. - liigh, l' eat Exchanger Rooms - East, West, bb. item 3.e, Reactor Vessel Water Level - Low Low, Level 2, cc. item 3.f. Main Steam Line Tunnel Ambient Temp. - High, dd, item 3,g, Main Steam Line Tunnel Delta Temp. High, ee. item 3.h, SLCS Initiation, ff, item 4.a, Reactor Core Isolation Coaling System Isolation, RCIC Steam Line Flow - High, gg. item 4.b, RCIC Steam Line Flow - High Timer, hh. item 4.c, RCIC Steam Supply Pressure - Low,

11. item 4.d, RCIC Turbine Exhaust Diaphragm Pressure - High, jj. item 4.e, RCIC Equipment Room Ambient Temperature - High, kk. item 4.f, RCIC Equipment Room Delta Temp. -

High,

11. item 4.g, Main Steam Line Tunnel Ambient Temp. - High, mm. item 4.h, Main Steam Line Tunnel Delta Temp. High, nn, item 4.1, Main Steam Line Tunnel Temp. Timer, oo. item 4.j, Drywell Pressure - High, pp. item 4.1, RHR/RCIC Steam Lino Flow - High, qq. item 4.m. RHR Heat Exchanger A,B Ambient Temperature - digh, rr. item 4.n, RHR Heat Exchanger A,B Delte. Temp. - High, ss. item 5.a, PHR Syntem Isolation, RRR Heat Exchanger Rooms A,B Ambient Temp. - lii gh ,

tt, item 5 b, RH2 Heat Exchanger Rooms A,B Delta Temp. - High, uu. item 5.c, Reactor Vessel Water level - Low, Level 3, vv. item 5.d, Reactor Vessel Watar Level - Low Low Low, Level 1, ww. item 5.e, Reactor Vessel (RRR Cut-in Permissive) Pressure -

High, xx. item 5.f.1, Drywell Pressure - High, RHR Test Line, and yy. item 5.f.2, Drywell Pressure - High, Fuel Pool Cooling.

5. The staggered test interval specified by footnote (a) to Technical Specification Table 4.3.2.1-1 is being increased from 31 days to 92 days .

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

7. An editorial change is being proposed to delete footnote "**"

associated with Surveillance Requirement 4.3.2.2 since this footnote was only applicable until the first refueling outage.

~- - - -. . - _ - . - . . .. . - - - - -

. Attachment 2

,' to U 601871' 1 LS-88-042 Page 18 of 33 Justification for Prqposed Chanres On August 29, 1985 the BWROC submitted Licensing Topical Report NEDC. 30851P, Supplement 2, " Technical Specification Improvement- Analysis for BVR Isolation Instrumentation Common- to RPS and ECCS Instrumentation,"

for NRC review. On June 27, 1989 the BWROC submitted Licensing Topical Report.NEDC 31677P, " Technical Specification Improveaent Analysis for BVR Isolation Actuation Instrumentation," for NRC review, The combination of the results from these two reports provides justification for the proposed changes identified as 1 through 6 above.

As stated in NEDC-30851P, Supplement 2, Technical Specification requirements for isolation instrumentation vere originally established largely on the basis of RPS and ECCS requirements. That is, the surveillance test intervals and allowable out of service times generally do not need to be morw stringent for isolation than for RPS or ECCS.

Even though isolation is a safety function, failure to isolate would not of itself result in an accident. The overall containment and reactor vessel isolation function is made up of several subfunctions, each of which must operate upon-demand during an accident. Failure of an isolation subfunction during an accident could potentially increase the offsite release risks.

The analysis presented in NEDC-30851P, Supplement 2 applies only to those CRVICS instruments which are common to the RPS or ECCS actuation instruments. Similar to the analyses discussed in Parts I and II of this submittal, fault trees were developed for each of the common isolation Trip Functions. These fault trees were then evaluated probabilistically to determine the irpact of the proposed changes on isolation unavailubility. As provided in NEDC-30851P, Supplement 2, the impact on the_ average isolation unavailability for the affected

= isolation tubfunctions due to the proposed changes was determined to be negligiole.(an increase of less than 1%) when combined with the ind'.vidual valve failure probabilities. The analyses demonstrate that tho individual valve failure probabilities dominate _the overall isolation failure probability.

By letter from Charles E..Rossi (NRC) to Donald N. Crace (BWROC) dated January 6, 1989, the NRC provided their Safety Evaluation Report of NEDC-30851P, Supplement-2. The NRC concluded in their Safety Evaluation Report that the mothods and results provided in NEDC-30851P, Supplement _

=

2 are-acceptable for implementation on a plant-specific-basis. However.

the NRC's Safety Evaluation' Report states that in order for a licensee to use the generic analyses'provided in NEDC-30851P, Supplement 2,.the licensee must confirm the applicability of the generic analyses to the plant and confirm that any increase in instrument drift due to the extended surveillence intervals is properly accounted for in the setpoint calculation methodology.

L With respect to the NRC Staff's concern about confirming the plant-specific applicability of NEDC-30851P, Supplement 2, the CPS CRVICS configuration was specifically modeled (identified as BWR-6= solid state) in NEDC 30851P, Supplement 2.

l Therefore, the generic results are directly applicable to CPS.

l l

l

. Attachment 2 to U 60'.871 LS-88-042 Page 19 of 33 With respt.ct to the NRC's concern about instrument drift, the instrument setpoint calculations for the CRVICS instrumentation at CPS include the effects of instrument drift over 18 months for all instrument loop componeres except for the analog trip modules. To address drift of the analog trip modules, IP reviewed the results of monthly calibration checks performed over a one-year period on the affected CRVICS analog trip modules which are common to RPS or ECCS. Review of these calibration checks showed that the quarterly drift is within the present calibration tolerances. As a result, IP has concluded that lengthening the CllANNEL FUNCTIONAL TEST interval and analog trip module calibration interval, as applicable, for the CRVICS instrumentation common to RPS or ECCS from monthly to quarterly will not result in excessivo drift relative to the current, established setpoints. In addition, a CllANNEL CllECK is required at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for those instruments with redundant channels. These routine CHANNEL CllECKS will help to identify excessive drift of the CRVICS instrumentation.

The analysis presented in NEDC-31677P applies to the remaining CRVICS Trip Functions (i.e., those CRVICS instrumnnts which are not common to RPS or ECCS actuation instrumentation). Similar to previous analyses discussed abovo, the analyses presented in NEDC 31677P are based upon fault trees (based upon the CPS design) which were evaluated to determine the impact of the proposed changes on the average isolation failure frequency. In this case, the average isolation failure frequency is defined as the product of the accident initiating event frequency (such as a pire break or high radiation event) and the probability of failure of the isolation function given a demand. The proposed changes were considered acceptable by the BWROC if the proposed changes resulted in less than a 10% increase in the average isolation failure frequency or if the average failure frequency was calculated to be Icss than 1.0x10'7/ year.

The results for the BWR-6 solid-state plent (which are directly applicable to CPS) demonstrate that these proposed changes only slightly increase the overall average isolation failure frequency for these instruments. AsidentifiedonTable5-3ofNEDC-3167fP, the calculated solation failure requency increased 4.4x10' / year from average 1.10x10' {/yearto1.54x10'{/ year.

By letter from Charles E. Rossi (NRC) to S. D. Floyd (BWROC) dated June 18, 1990, the NRC provided their Safety Evaluation Report of NEDC-31677P. The NRC concluded in their Safety Evaluation Report that the methodology and acceptance criteria provided in NEDC-31677P are acceptable for implementation on a plant-specific basis, llowever, the NRC's Safety Evaluation Report states that in order for a licensee to use the generic analyses presented in NEDC 31677P, the licensee must confirm the applicability of the generic analyses to the plant and confirm that any increase in instrument drift due to the extended surveillance intervals is properly accounted for in the setpoint calculation methodology.

As identified above, the CPS CRVICS configuration was specifically modeled (identified as BWR-6 solid-state) in NEDC-316/7P. Therefore, the generic results of NEDC-31677P are directly applicabic to CPS.

, Attachment 2 to U 601871 LS-88 042 Pa6e 20 of 33 With respect to the NRC's concern about instrument drift, the instrument setpoint ca*culations for the CRVICS instrumentation at CPS include the effects of instrument drift over 18 months for all instrument loop components except for the analog trip nodules. To address drift of the analog trip modules, IP reviewed the results of monthly calibration checks performed over a one-year period on the affected CRVICS analog trip modules which are not common to RPS or ECCS. Review of these calibration checks showed that the quarterly drift is within the present calibration tolerances. As a result, IP has concluded that lengthening the CRANNEL FUNCTIONAL TEST interval and analog trip module calibretion intetval, as spplicable, for the CRVICS instrumentation not common to RPS or ECCS from monthly to quarterly will not result in excessive drift relative to the current, established setpoints. In addition, a CRANNEL CHECK is required at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for those instruments with redundant channels. These routine CRANNEL CHECKS will help to identify --

excessive drift of the CRVICS instrumentation, it should be noted that the format of the Action Statement's for the CRVICS instrumentation proposed in this request differs from that provided in NEDC-30851P, Supplement 2 and NEDC-31677P and from that provided in the NRC Safety Evaluation Reports for these reports. The above documents recommended replacir.g Action b (equivalent to CPS Action c.1) with an Action Statement which incorporated footnote "**" and recommended separate requirements for those instruments that are common _

to RPS. This format vos recommended because the BVR 6 relay type plants' requirements should be based on whether the channel can be placed in the tripped condition without resulting in an actuation while recognizing the shorter A0T of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for the relay plants' RPS instrumentation. Therefore, when placing the inoperable channel in the tripped condition would cause an isolation, the above documents recommended a repair A0T of six hours (equivalent to the surveillance AOT). When placing the inoperable channel in the tripped condition would not cause an isolation, the above documents recommended that the inoperable channel be placed in the tripped condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> -

for instruments common to RPS and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for instruments common to ECCS.

As previously identified, the CPS RPS logic design is solid-state. This solid state RPS logic is combined in a two-out-of-four arrangement such that a trip of any two of the four RPS channels will result in a reactor scram. Because of this redundancy, one RPS channel is currently allowed to be inoperable for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. As proposed in Part I of this submittal, two RPS channels may be inoperable for up to six hours.

Because one RPS channel may be inoperable for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> (vs. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for BWR-6 relay-type plants), there is no need to specify a shorter A0T for CRVICS instruments common to RPS. Hence, an Action Statement that provides a separate A0T for CRVICS instruments common to RPS is not required for CPS. The proposed Action c.1 will permit, for all CRVICS instruments (including those common to RPS), a maximum A0T of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (vs. 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> per Technical Specification 3.3.1, " Reactor Protection System Instrumentation."). Additionally, footnote "**" has been retained, but the repair A0T has been increase; from two hours to six hours (consistent with the proposed surveillance A0T) . As a result, proposed Action c.1 is consistent with the model Technical

. Attachment 2 to U-601871 LS-88-042 l Page 21 of 33 Specifications provided in the BVROG reports as applied to the BVR 6 solid-state design.

l One additional area that was not specifically identified in the model l Technical Specifications provided in the BVROG reports relates to CPS Action b.for the CRVICS Main Steam Line Isolation Trip Punctions. These Trip Functions utilize logic which is identical to the RPS. As such, these requirements should be identical to the Action Statements of Technical Specification 3.3.1. Accordingly, Action b.2 has been revised to match the changes proposed in Part I of this submittal.

Ensis For No Sir _nificant llazards Consideration In accordance with 10CFR50.92, a proposed change to the operating license (Technical Specifications) involves no significant hazards -

considerations if operation of the facility in accordance with the proposed change would not: (1) involve a significant increase in the probability or consequences of any accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety. The proposed CRVICS instrumentation Technical Specification caanges are evaluated against each of these criteria below.

(1) These' proposed changes do not involve a change to the plant design or operation, only to the allowable out of-service time (A0T) and ,

frequency at which testing of the CRVICS instrumentation is performed. Failure of the CRVICS instrumentation itself cannot create an accident. As a result, these proposed changes cannot increase the probability of any accident previously evaluated.

As identified in NEDC-30851P, Supplement 2, the proposed changes to the requirements for the CRVICS instruments which are common to RPS and ECCS have a negligible (less than 1%) impact on the average isolation unavailability when combined with the individual valve failure probability. The analyses demonstrate that the individual valve failure probabilities dominate the overall isolation failure probability. As identified in NRC Staff's Safety Evaluation Report of NEDC-30851P, Supplement 2, these proposed changes would have a very small impact on plant risk. As a result, overall plant safety is not reduced by these proposed changes.

As identified in NEDC-31677P, the proposed changes to the not common to RPS or ECCS requirementsforCRVICSinstrumentatiog/yearintheaverage result in a small increase of 4.40x10' isolation failure frequency. As identified in the NRC Staff's Safety Evaluation Report of NEDC-31677P, the NRC agreed that these proposed changes are acceptable because the failure frequency impact is minimal As a result, overall plant safety is not reduced by these proposed changes.

The small increase in the average failure frequency due to the proposed changes to the CRVICS instrumentation requirements is offset by safety benefits such as a reduction in the number of l

l

Attachment 2 to'U 601871 LS-88-042 Page 22 of 33 inadvertent test-induced scrnis and enaineered safety feature actuations, a reduction in wear due to excessive test cycling, and better optimization of plant personnel esources. Therefore, the proposed changes do not represent a significant increase in the consequences of any accident previously evaluated'.

(2) These proposed changes do not resu4t in any change to the plant design or operation, only to the A0T and frequency at which testing of the CRVICS instrumentation is performed. Since failure of the CRVICS instrumentation itself cannot create an accident, these proposed changes can at most affect only accidents which have been previously evaluated. Therefore, these proposed changes cannot create the possibility of a new or different kind of accident from any accident previously evaluated.

(3) As identified above, these proposed changes to the requirements for CRVICS instruments common to RPS or ECCS have a negligible impact on the average isolation unavailability when combined with the individual valve failure probability. The analyses demonstrate that the individual valve failure probabilities dominate the overall isolation failure probability. The proposed changes to the requirements for CRVICS instruments not common to n failure frequency RPSorECCSincreasegheiraverageisolatig/yearto approxim9tely4.4x10"/ year (from1.10x10' 1.54x10' / year). This increase is negligible.

The small increases in average CRVICS instrumentation failure frequency are offset by safety benefits such as a reduction in the number of inadvertent test-induced scrams and engineered safety feature actuations, a reduction in wear due to excessive equipment test cycling, and better optimization of plant personnel resources. As a result, the NRC Staff's Safety Evaluation Reports for these BWROG reports concluded that these proposed changes would have a very small impact on plant risk. In addition, IP has confirmed that the proposed changes to the functional test int vals will not result in excessive instrument drift relative to the current, established setpoints. Therefore, these prnpo'ed s changes do not result in a significant reduction in a margin of safety.

Based upon the foregoing, IP concludes that these proposed charges do not involve a significant hazards consideration.

l

.._ Attachment 2 l

• '_ to U 601871-LS-88 042 Page 23 of 33

~_

Part V - Qthe;.r_TnhMaid.Aul.USE1Rn.InGrm.iluh Description of Proposed Chanr.es 1

In accordance with 10CPR50.90, the following changes are proposed:

1. Irchnical Specification 3/4.3.4.1. "ATWS Recirculation Pumn Trin System Instrumentation"

a. The surveillance allowable out of-service - time ( A0T) of ,

footnote (a):to_ Technical Specification Table 3.3.4.1-1, i "ATWS Recirculation Pump Trip System Instrumentation," is being increased from two hours to six hours,

b. The CHANNEL FUNCTIONAL TEST interval specified on Technical .

Specification Table 4.3.4.1-1, "ATWS Recirculation Pump Trip '

Actuation Instrumentation Surveillance Requirements," is being increased from monthly (M) to quarterly (Q) for the following Trip Functione:

(1)_ item 1, Reactor Vessel Water Level - Low low, Level 2, and (ii) item 2, Reactor Vessel- Pressure - liigh. 7

c. The trip unit calibration interval specified by footnote "*"

to Technical Specification Table 4.3.4.1-1 is being increased from 31-days to 92 days, ,

2. Technical Specification 3/4.3.4,2. "End-of-Cvele-Recirculation

,. Pump Trio System Instrumentation"

a. - The repair. A0T of ' Action e is being -increased from one hour to six hours,

b. The surveillance A0T of_ footnote (a) to Technical Specification Table 3.3,4.2-1, "End-of Cyclu Recirculation Pump Trip System Instrumentation,"_is being increased from two hours to six hours.

c. The CHANNEL FUNCTIONAL TEST interval specified on Technical Specification Table 4.3.4.2-1, "End-of-Cycle Recirculation:

Pump Trip System surveillance Requirements," is being increased from monthly-(M) to quarterly (Q) for tha following Trip Functions:

r (i) item 1, Turbine _Stop_ Valve- Closure, and (ii) item 2, Turbine Control Valve Fast Closure,

d. An editorial _ change-is being proposed to delete the " Total Number af' Channels" and the." Channels to Trip" columns of -

Technical Specification Table 3.3.4.2-1 and revise the-Minimum _ OPERABLE Channels per Trip Function requirements from "3" to "4". Since this logic is arranged in a two out-

- of-four scheme and the Action Statements address channel p inoperability based on four channels, this proposed change l-i l-E __. _ - _

Attachment 2 to U 601871 LS-S8-042

. Page 24 of 33 is being made to make Table 3.3.4.2-1 match the current Action Stateme..ts. This proposed change does not reduce the number of channels required to be OPERABLE.

c, An editorial change is being proposed to delete footnote "*"

associated with Surveillance Requir +ent 4.3.4.2.2 since this footnote was only applicable u.. 'l the first refueling outage.

3. Itchnical Specification 3/4.3.5. " Reactor Core Isolation Cooline.

System Actuation Iqstrumentation"

a. The repair A0Ts of Tecnnical Specification Table 3.3.5-1,

" Reactor Core Isolation Cooling System Actuation Instrumentation," Actions 50.a and 52 are being increased -

from one hour to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; Action 53 is being increased from eight hours to 24 hourr; and Action $1 is being identified as 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

b. The surveillance A0T of footnote (a) to Technical Specification Table 3.3.5-1 is being increased from two hours to six hours.

c. The CilANNEb FUNCTIONAL TEST interval specified on Technical Specification Table 4.3.5.1-1, " Reactor Ccre Isolation Cooling System Actuatiou Instrumentation Surveillance Requirements," is being increased from monthly :M) to quarterly (Q) for the following Functional Units:

(i) item a, Reactor Vessel Ucter Level - Low Low, Level 2, (ii) item b, keactor Vessel Water Level - High, Level 8, (iii) item c, RCIC Storage Tank Level - Low, and (iv) item d, Suppression Pool Water Level - High.

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

e. An editorial change is being proposed to delete footnote "*"

associated with Surveillance Requirement 4.3.5.2 since this footnote was only applicable until the first refueling outage

4. Technical Specification 3/4.3.6. " Control Rod Block Instrumentation"

a. Action 64 is being added to Technical Specification Table 3.3.6-1, " Control Rod Block Instrumentation," to allow control rod block instrumentation channels to be inoperable for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> rather than one 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:

, Attaciunent 2 to U 601871 LS-88 042 -

Page 25 of 33

, (i) item 5.a. Scram Discharge Volume, Water Level High, and (ii) item 6.a Reactor Coolant System Recirculation Flow, Upscale.

5. Iechnical Specification 3/4.3.7.1. " Radiation Monitorinn Instrigertation"

a. The repair AOT of Technical Specification Tabic 3.3.7.1-1,

" Radiation Monitoring Instrumentation," Action 70.a is being increased from one hour to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. The CHANNEL FUNCTIONAL TEST interval specifled on Technical Specification Table 4.3.7.1-1, " Radiation Monitoring Instrumentation Surveillance Requirements," is being increased from monthly (M) to quarterly (Q) for item 1, Main Control Room Air Intake Radiation Monitor.

6. Technical Specification 3/4.3.9. " Plant Systems Actuation Instrumentation" -

a. The repair A0T of Action b.1 and Technical Specification Table 3.3.9-1, " Plant Systems Actuation Instrumentation,"

Action 50.a in being increased from one hour to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and a repair A0T of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is being added to Technical Specification Table 3.3.9-1 Action 51.

b. The surveillance A0T of footnote "*" to Technical Specification Table 3.3.9-1 is being increa.ed from two hours to six hours,

c. The CHANNEL FUNCTIONAL TEST interval specified on Technical Specification Table 4.3.9.1-1, " Plant Sjstems Actuation Instrumentation Surveillance Re32irements," is being increased from monthly (M) to quarterly (Q) for the following Trip Functions:

(1) item 1.a Containment Spray System, Drywell Pressure -

High, (11) item 1.b, Centainment Pressure - High, (iii) item 1.c, Reactor Vessel Water Level - Low Low Low, Level 1, (iv) item 1.d, Timers, (v) item 2.a. Feedwater System / Main Turbine Trip System, Reactor Vessel Water Level - High, Level 8, (vi) item 3.a. Suppression Pool Makeup, Drywell Pressure -

High, (vii) item 3.b. Reactor Vessel Water Level - Low Low Low, Level 1, (viii) item 3.c, Suppression Pool Water Level - Low Low, and (ix) item 3.d, Suppression Pool Makeup Timer.

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

, Attachment 2 to U 6018/l LS 88 042 page 26 of 33

e. The analog comparator unit calibration interval specified by footnote (b) to Technical Specification Table 4.3.9.1 1 is being increased from 31 days to 92 days.

i An editorial change in being proposed to correct the spoiling of the word "comparator' in footnote (b) to Technical Specification Table 4.2.9.1 1.

,. An editorial change is being proposed to Action a to waka it consistent with the wording utilized in similar Action Statements of other Technical Specifications,

h. An editorial change ir also being proposed to delete footnote "*" associated with Surveillance Requirement 4.3.9 2 ance this footnote was only applicabic until the first refueling outage.

6

7. IfXlmical Speci ficaljon 3/4.4.2.1, " Safety / Relief Valver"

a. The surveillance A0T of footnote "* is being increased from two hours to six hours.

h. The CllANNEL PUNCT10NAL TEST interval of Surveillance Requirement 4.4.2.1.2.a is boing increased from 31 days to 92 days.

c. An e:iitorial change is being proposed to delete footnote

"**" associated with Surveillance Requirement 4.4.2.1.2.b since this footnote was only appilcabic until the first refueling outage.

8. I ;aital Specification 3/4.4.2.2. "Sofety/ Relief Valves Low-Low Set Fgcc.Linn"

6. The surveillance A0T of footnote "*" is being increased from two hours to six hours,

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

c. An (; '+oria; change is being proposed to delete footnote "6" as: .iated with Surveillance Requirement 4.4.2.2.b since th;- Tootnote was only applicable until the first refueling outa6e-Justific3 tion for Proposed Channt]

On February 19. 1991 the PVROC submitted Licensing Topical Report GENE.

770 06 1, " Bases for Changes to Surveillance Test Intervals and Allowed Out-of' Service Times for Selected Instrumentation Technical Specifications," for NRC teview. This report provides the justification for the proposed changes identified above (except those identified as editorial changes). Although CENE 770-06 1 has not yet been approved by l

J A_ -- - - - - - - - --- - _

c

• . Attachment 2 to U.6011tyl LS 88 042 j Page 27 of 33 the NRC Staff. IP is requesting the Techaical Specification changes identified in that report (as described above) at this time in order to  !

provide a cornplete request with respect to the instrumentation reliability based improvements. If the proposed changes in Part V of  ;

this request are not approved, corne of the 1:nprovement.s in Parts I ,

through IV of this request would not be able to be implemented. This is because these instruments perform tuultiple functions which are addressed '

by separate Technical Specifications and hence, are addressed by I separate Topical Reports,  ;

As ts ed in CENE 770 06 1, the primary purpose for requesting these '

changes is to ensure consistency with the changes proposed for the RPS, l ECCS actuation instrweentation and CRVICS sctuation instrweentation, The instrweentation affected by the proposed changes in Part V cf this request affect either- the same or sit 11ar instrumentation addrer ' in >

Parts I through IV. The primary difference fs the safety fune perfortned by the instrumentation.

As also noted in CENE.770 06-L a detatied onalysis of those proposed l changes that are associated with instrumentation that is common to  !

previously analyzed-instrumentation was not performed since the analyses -

discussed in Parts I through IV bound them. The remahing proposed i changes involve instruments which are of a similar type to the instruraents included in the analyses discussed in Parts 1 through IV.

Existing redundancy of'this instrumentation is either more extensive or comparable to the reduridancy of the instrwnents -discussed in Parts I '

through IV. Further, analyses have generally shown that the most.

significant contributor to safety function failure probability is associated with the actuated device (such as valves) rather than associated with the actuation instrumentation. Therefore, the analyses discussed in Parts I through IV of this request can be used to justify '

the proposed changes identified in its part.

• As discussed in Parts I through IV of this request, any expected increast in the probability of function failure as a result of these proposeo changes will be offset by safety benefits such as a reduction -

in the number of inadvertent test induced scrams and engineered safety feature actuations, a reduction in wear due to excessive equipment test cycling, and better optirnization of plant personnel resources. As a t result, these proposed changes do not result in a degradation to overall plant safety.

The basis for IP's deterinination that each of these proposed changes are <

bounded by the analyses discussed in Parts I through IV of this request is discussed below for each of the affected systeins, 1.

• fechnical Swelfication 3/4.3.4.1. "ATVS Recirculation Pumn Trin Systems Instrumentation" The ATVS RPT instrumentation is part of the'initigation system : hat- '

initiates in the unlikely event of a. scram failure, The trip function is initiated by either high reactor pressure or low low

l. reactor water _ level (Level 2), The ATWS.RPT logic for CPS is two-out of+two channels per trip system for each Trip Function. Each of the two trip systeins initiates a trip of both recirculation L

l,- -

Attochment 2

' to U 601871 IS-88-042 Page 28 of 33 ptumps . The effect of the proposed changes to the A1VS-RPT instrusentation requireinents on the reactivity shutdown failure ible based on the low average RPS failure frequency is neglig/ year from NEDC 30851P, page 6 11) frequency (2.3x10* and the sinall change ion unavailability due to in overall(less the proposed changes ATWS RPT than funeg/dernand calculated from In10' failure rates of siinilar instruments as given in Appendix B and C of NEDC 30851P).

2. Technten1 Specificat ion 3/4. 3.4.2. "End-of Cvele % eiten}_at ion hurn Trip Sys t em Inst rument at ion" The EOC PPT is initiated by signals and inatruirentation cornmon to the RPS (turbine stop valve closure and turbine control valve low hydraulic pressure). The proposed changes for this instrumentation were evaluated in NEDC 30851P for the RPS function. Although the EOC RPT trip functions were not explicitly ident.ified in NEDC 30851P, these proposed changes can be considered bounded by that analysis. The basis for this conclusion is similar to the basis established in NEDC 30851P, Suppleinent 2 for the control rod block instrumentation coimon to the RPS. That is, although failure of the EOC-RPT trip function could potentially lead to exceeding the Minientua Critical Power Ratio (MCPR) limit (similar to the consequences of an usunitigat ed rod withdrawal etror), the slight increase in risk of an MCPR violation due to the proposed EOC-RPT changes is offset by the safety benefits associated with the proposed chances for the RPS instrumentation.

3. Technical Specificat.lon 3/4.3.5. " Reactor Core irdL.llon Cooling System Actuation instrumentation" The proposed changes to the RCIC system actuation instrumentation were evaluated in the BWROC analysis of ECCS actuation instrtunentation (NEDC 30936P (Part 2)). The RCIC fault. tree models and input data were developed for the CPS design (BWR-6 solid state). In NEDC 30936P (Part 2), t he water injection function failure frequency was analyzed as a function of the STis and ADTs for the ECCS (including RCIC) ac wation instrumentation.

The RCIC actuation instrumentation turve tilance test interval (STI) was changed from 1 to 3 months and the associated A0T was changed frein 1 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for repair and from 2 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for test. The analysis results are summarized in NEDC-30936P (Part 2); however, model Technical Specification changes for the RClc actuation instrumentation were not specifically included in NEDC-30936P (Part 2).*

• Model Technical Specification changes f or the RCic actuation instrumentation were thus later provided in CENE-770-06-1.

. Attachment 2 to U.601871 LS 88 042 Page 29 of 33 An analysis was conducted to demonstrate the specific effect of individual changes to the RCIC actuation instrumentation STis on the overall average water injection function unavailability. As noted above, the analysis was performed using the models and input data developed and documented in NEDC 30936P (Parts 1 and 2). In order to determine the specific effect of the STI change on the RCIC actuation instrumentation, the RCIC actuation instrumentation STI was held constant (i.e., STI - one toonth) while the STI f or other ECCS actuation instrwoentation was changed to three months.

This calculation demonutrated that thece is a very small change in the calculated average water injection function unavailability (less than in) for this caso when cotepared with the results of NEDC-30936P (Part 2). The NEDC 30936P (Part 2) analysis results indicated that the effect of A0T changes is significantly less than STI changes. On this basis, a siinilar negitgible change in average water .'jection function unnva11 ability can be expected when the RCIC a tuation instrumentation AUTs (one hour repair and two hours test) e e held constant. Therefore, it can be concluded that the STI ano .s0T changes to the RCIC actuation instrumentation are justified based on the small ef fect on the calculated average water injection function unavailability and consister.cv with comparable changes to the actuation instrumentation for the ECCS subsystems.

4. Technien1 Specification 3/4.3.6. " Cont rol Red Block J rm t rument e t t on" NEDC-30851P, Supplement 2 provided the bases for changing the ST1s for the control rod block instrumentation from one month to three months. Although the above changes to the repair and tout A0Ts were not explicitly identified in NEDC 30851P, Supplement 2, the same bases used for changing the ST1s applies to the A0T changes.

The reason for this is because analyses indicate that the effect of A0T changes is significantly less than the effect of STI changes. The proposed changes to the A0Ts for the control rod block instrumentation are therefore supported by the basis provided in NEDC 30851P, Supplement 2.

5. Technical Specificpt ion 3/4. 3,7.1. "Rarjj,at ion Moni t oring Instrumentation" The main control room ventilation systein is provided with radiation monitors to monitor radiation Icvels at the two outside minimum air intakes. Upon detecting a high radiation signal, the main control room ventilation system is automatically placed into the filtration mode. A Division I and a Division 11 radiation monitor is provided at each air intake. The radiation monitor outputs are combined in a one-out of-two-twice logic to actuate the automatic filtration mode. This instrumentation arrangement is similar to the CRVICS radiation monitoring instrumentation.

Therefore, the analysis of the isolation actuation instrumentation provided in NEDC 31677P supports similar STI and A0T changes to the Technical Specifications for the main control room ventilation intake radiation monitors.

Attachment 2 to U 601871 LS 88 042

. Page 30 of 33 i

l 6. Technical Speci ficat iott.344. 3. 9. "Plent 5.ys t ems Ac t uathn 1Dstrumentatign This Technical Specti1 cation addresses the requirements for those itntruroents that provide autornatic ac' uation of the contaitunent spray system, feedwater/ main turbino trip system, and the suppression pool makeup system. Each af these systems are discussed separately below.

a. Egg nment Spray System The contaitunnnt spray system actuation instrumentation contains instrumentation comrnon to the ECCS actuation instrumentation. In addition, the actuation function performed (i.e., closing and opening selected valves) is similar to the function performed by the isolation and ECCS actuation instrumentation. The dominant contributor to the unavailability for this type of function is valve unavailability. Therefore, the analyses of isolation actuation instrumentation provided in NEDC 30851p, Supplement 2 and NEDC 31677P support similar STI and A0T changes to the containment spray system instrumentation,

b. fredwater System / Main Turbine...Tripl vstem The BWR 6 plant design incorporates a direct scram from hi S h reactor vessel water level (hevel 8) trip instrumentation (included in the RPS instrumentation). The bases for changes to the STIs and A0Ts for the reactor vessel water level 8 trip instrumentation associated with the feedwater system / main turbine trip system are therefore bounded by the changes to the RPS reactor vessel water Level 8 trip instrumentation provided in NEDC 30851P.

c. Suppression Pool Makeun The same bases given for the containment spray system instrumentation applies for the suppression pool makeup system instrumentation. The suppression pool makeup system instrmentation contains instrumentation which is common sa the E.,CS actuation instrumentation. In addition, the actuation function performed (i.e., opening selected valves) is similar to the function performed by the isolation and ECCS actuation instrumentation. The dominant contributor to the unavailability for this type of function is valve unavailability. Therefore, the analyses of isolation actuation instrumentation provided in FEDC-30851P, Supplement 2 and NEDC 31677P support similar STI and A0T changes to the suppression pool makeup system instrumentation.

7. Technical Specification 3/4.2.4.2.1. "SafetylFelief Valves" For CPS, fi.ve of the 16 safety / relief valves (SRVs) are required to open in the relief mode (actuated by a pressure transmitter)

M/%NN 6 Attachment 2 to U.601871 LS 88 042 page 31 of 33 and six are required te open in the safety mode (actuating against spring pressure) to prevent reactor vessel overpressurization.

SRV safety mode actuation is diverse from the relief mode actuation. The relief function of the SRVs is performed bf three separate sets of logic. Each logic set is actuated by one of two two out of-two reactor steam dome pressure logic combinations.

The first logic group controls the relief function for onc valve, the second logic group controls eight valves, and the third logic group controls seven valves. If a relief function logic group should fail (which requires at least two channel failures),

overpressure protection can be provided by the remaining relief logic groups in combination with SRV actuations in the safety mode.

Based on the level of redundancy, unavailability of the relief valve pressure actuation function is a small contributor to the overall SRV function unavailability. Changes to the STI and A0T for the SRV pressure actuation instrumentation will therefore have an insignificant effect on the probability of failure to prevent reactor overpressurization. 1hese STI and A0T changes will also be consistent with STI and A0T changes to similar instrumentation in the ECCS and isolation actuacion systems.

8. Technical Speciftention 3/6.4.2.2. "Safetv/ Relief Valves low-low Set Function" The Low Low Set (LLS) logic for CPS consists of three individual LLS circuit groups ahich control five LLS SRVs. This logic is designed so that no more than one SRV reopens following a reactor vessel isolation event, ensuring that the containment design basis 4

is met. After an LLS SRV initially opens in the relief mode, the associated LLS logic is activated and the SRV's closing setpoint is lowered such that the SRV stays open longet than without LLS.

Two of the LLS circuit groups each control an individual SRV.

These two logic circuit groups also lower the SRV's reopening setpoint such that the SRV will open prior to activating additional SRVs in the relief mode. The third logic circuit group controls a group of three LLS SRVs and only lowers their closure setpoints. The LLS function can normally be performed by either of the first two LLS logic groups.

Because energization of either SRV solenoid pilot valve results in opening the SRV, both solenoid pilot valves must be de-energized for the SRV to close. Opening of the first two LLS logic groups is accomplished by actuation of one of the two two out of two SRV relief mode logic trains. Subsequent closure and reopening of these two LLS logic groups is accomplished by actuation of a one-out-of onc logic for each solenoid pilot valve. The third LLS logic group opens upon actuation of one of two two out of two logic trains and recloses upon deactivation of both two-out of two logic trains.

Although the LLS logic has an important safety function, its function is not as critical to overall plant safety as the water injectic e or isolation functions. Therefore, changes to the STIs

l Attachment 2 to U 601871 LS.88 042 Page 32 of 33 and A0Ts for the LLS pressure actuation instrumentation will have less risk impact on the overall plant safety than ECCS and isolation actuation STI and A0T changes. Existing analyses of ECCS and isolation actuation instrument ation STI and A0T changes can be applied to the LLS logic based on the use of the same or similar type of components (i.e., relays, transmitters, trip units, etc.), designed redundancy, and safety significance of the LLS logic. The extensive redundancy in the LLS circuit logic is comparable with the lob i c redundancy in the ECCS and isolation actuation instrumentation. Based ca this redundancy, similarlty of components, and safety function sl 6nificance of the LLS logic, it can be concluded that the ef fect of changes for the LLS logic STIs and A0Ts is bounded by the basis established for similar STI and A0T changes for the ECCS and isolation actuation instrumentation.

With respect to instrument drift for all of the instrumentation addressed in this part, the instrument setpoint calculations for these instruments include the effects of instrument drift over 18 months for all instrument loop components except for the trip units. To addrens drift of the trip units, IP reviewed the results of monthly calibration checks performed over a one-year period on the affected trip units.

Review of those calibration checks showed that the quarterly drift. is within the present calibration tolerances. As a result, Ip has concluded that lengthening the CllANNEL TUNCTIONAL TEST interval and t rip unit calibration interval, as applicable, for the affected instrementation from monthly to quarterly will not result in excessive drift relative to the current, established setpoints. In addition, a CilANNEL CllECK is required at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for those instruments with redundant channels. These routine CilANNEL CilECKS will help to identify excessive drift of the instrumentation affected by these proposed changes.

Basis for No Sir.nificant llazarde Consideration In accordance with 10CFR50.92, a proposed change to the operating license (Technical Specifications) involves no significant hazards considerations if operation of the facility in accordance with the proposed change would not: (1) involve a significant increase in the probability or consequences of any accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previous 13 evaluated, or (3) involve a significant reduction in a margin of safety. The proposed Technical Specification changes are evaluated against each of these criteria below.

(1) These proposed changes do not involve a change to the plant design or opera' tion, only to the allowable out-of service time (A0T) and frequency at which testing of the associated instrumentation is performed. These instruments are designed to mitigate the consequences of previously analyzed accidents. Failure of these instrumerts cannot increase, and is independent of, the probability of occurrence of such accidents. As a result, these proposed changes cannot increase the probability of any accident previously evaluated. As identified in CENE-770-06 1, although not specifically analyzed, these proposed changes are bounded by

, At.techment. 2

• to U.601871 LS 88 042  ;

Page 33 of 33 ,

the results of the analyses discussed in parts I through IV of this request. Such analyses have shown that the safety function ~

failure frequency is not significantly impacted by siellar ,

proposed changes. In addition, any increase in the probability of '

failure-of these instruments to perform their required functions i would be offset ty safety benefits such as a reduction in the '

number of inadvertent test induced scrams and engineered safety {

festure actuations, a reduction in wear due to excessive equipment i test cycling, and better optimization of plant personnel resources. As a result, these proposed changes should reduce i overall plant risk. Therefore, these proposed changes do not '

result in.a significant increase in the probability or the  !

consequences of any accid:nt previously evaluated.  ;

r (2) These proposed changes do not result in any change to the plant

. design or operation, only to the A0T and frequency at which l

testing of the associated instrumentation is performed. As a result, these proposed chan6es can at most affect. only accidents  !.

which have been previously evaluated. Therefore, these proposed changes cannot create the possibility.of a new or of h*erent kind '

of accident from any accident previously evaluated.

(3) As identified in CENE.770 06 1, although not specifically  ;

analyzed, these proposed changes are bounded by the results of the. .,~

analyses discussed in Parts I through IV of this request. Such analyses have shown that the safety function failure frequency is '

not significantly impacted by similar proposed changes. In addition, any increase.in the prob Aility of failure or these instruments to perform their required functions wouid be offset by safety benefits such as.a reduction in the number of inadvertent I

teat induced scrams and engineered safety feature actuations, a redaction in wear due to excessive equipment test cycling, and better optimization of plant personnel resources. As a result,

' these proposed chan6es will reduce overall plant risk. .In addition,-IP has confirmed that the proposed changes'to the

• functional test intervals will not result in excessive instrument

' drift relative to the current, established setpoints. Therefore, these_ proposed changes.do not involve a significant reduction in a margin of safety.

Based upon the foregoing, IP har, concluded that these proposed chan6eS

(. ..do not involvo a significant hazards consideration.

1

),

l.

l t

(. b

, - . . _ _ _,..,_,.-m m.m..,.,_ , -,..,, J_,...., % ,._ ,_.,m... __ ...-,m .__.g_. p , . . . , , .,-,,~,f._