ML20101L714

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Proposed Tech Specs,Revising APRM Flow Transmitter Calibr, Definition of Operable,Automatic Dispatch Sys Logic Mods & Organization Change
ML20101L714
Person / Time
Site: Cooper Entergy icon.png
Issue date: 12/20/1984
From:
NEBRASKA PUBLIC POWER DISTRICT
To:
Shared Package
ML20101L709 List:
References
NUDOCS 8501020350
Download: ML20101L714 (13)
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Attachment 1 Revised Technical Specifications for APRM Flow Transmitter Calibration Revised Pages: 41 81 Recently, it was determined that the actual operation of the APRM flow biased scram and rod block systems is inconsistent with the description given in the CNS Technical Specifications. Technical Specifications, Section 4.1, Bases, Indicates that while calibrating the APRM Flow Biasing Network, ". . . a zero flow signal will be sent to half the APRM's resulting in a half scram and rod block condition." In actuality, when in calibrate mode, each reactor recirculation flow unit sends a full flow signal to half of the APRM's, producing a rod block, but not a half scram. This condition was reported in LER 50-298-84-09.

An analysis that was conducted concluded the system reliability was unaffected and that a substantial margin from fuel damage was provided by the 120% high flux scram which is in effect during the calibration. The CNS Updated Safety Analysis Report relies only upon the 120% high flux scram and takes no credit for the APRM flow bias scram. The analysis further concluded there is no loss of safety function to the extent that there is a reduction in the degree of protection provided public health and safety.

Nebraska Public Power District requests a revision to the Technical Specifications to reflect the actual operation of the APRM flow biased scram and rod block systems. This revision is judged to involve no significant hazards based on the following:

Evaluation of this Revision with Respect to 10CFR50.92

1. Does the proposed license amendment involve a significant increase in the probabil!.ty or consequences of an accident previously evaluated?

i Evaluation:

Analysis shows that scram and rod block system reliability is unaffected and that the APRM flow bias scram which is affected only during flow i transmitter calibration is not taken credit for in the CNS Safety l Analysis. The f requency of occurrence, during the calibration, of a plant transient which would be terminated by the APRM flow biased scram has been conservatively estimated at once per 500 year and can therefore be classified as an infrequent event. Again, it is noted the 120% APRM scram trip setting is operable at all times and is the trip setting used in the safety analysis. From the above, it is judged the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. -

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2. Does the proposed license amendment create tb, possibility of a new or different kind of accident from any accident previously evaluated?

Evaluation:

Because the 120% high flux APRM scram is always in effect and was used in the Safety Analysis, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

8501020350 841220 PDR ADOCK 05000298 P PDR

_ 3. Does the proposed amendment involve a significant reduction in a margin of safety?

Evaluation:

Because the 120% high flux APRM scram is always in effect and has been shown by the Safety Analysis to provide a substantial margin to fuel damage for all abnormal operation transients analyze, the ' proposed i amendment does not involve a significant reduction in a margin of safety.

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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.1 BASES (Cont.d) 4.1 BASES (Cont.d) ence paragraph VII.S.7 FSAR). Thus full scale flow signal will be sent l the IRM System is not required in to half of the APRM's resulting in the "Run" mode. The APRM's cover a rod block condition. Thus, if the l only the power range. The IRM's calibration were performed curing and APRM's provide adequate coverage operation, flux shaping would not in the startup and intermediate range, be possible. Based on experience at other generating stations, drift The requirement to have the scram of instruments, such as those in functions indicated in Table 3.1.1 the Flow Biasing Network, is not operable in the Refuel mode assures significant.

that shifting to the Refuel mode during reactor power operation dces Group (C) devices are active only not diminish the protection provided during a given portion of the by the reactor protection system. operational cycle. For example, the IRM is active during startup Turbine stop valve scram occurs at and inactive during full power 10% of valve closarc. Below 233 psig operation. Thus, thc only test turbine first stage pressure (30% of that is meaningful is the one rated), the scram signal due to tur- Performed just prior to shutdown bine stop valve closure is bypassed or startup; 1.e., the tests that because the flux and pressure scrams are performed just prior to use are adequate to protect the reactor. of the instrument.

Turbine control valves fast closure Calibration frequency of the instru-initiates a scram based on pressure ment channel is divided into two switches sensing Electro-Hydraulic groups. These are as follows:

Control (LilC) system oil pressure.

The switches are located on the 1. Passive type Judicating devices Control Valve Emergency Irip oil that can be compared with like header, and detects the loss of units on a continuous basis.

oil to hold the valves open.

2. Vacuum tube or semi-conductor This scram signal is also bypassed devices and detectors that when turbine first stage pressure drif t or lose sensitivity.

is less than 233 psig.

Experience with passive type instn -

The requirements that the IRM's be in- ments in generating stations and sub-serted in the core when the APRM's ren stations indicates that the specified 2.5 indicated on the scale in the calibrations are adequate. For those Startup and Refuel modes assures that devices which employ amplifiers, etc.,

drift specifications call for drift to be less that 0.4%/ month; i.e., in the period of a month a maximum drift of 0.47, could occur, thus providing for adequate margin.

NOTES FOR TABLES 4.2.A THROUGH 4.2.F

1. Initia11g once every month until exposure (M as defined on Figure 4.1.1) is 2.0 X 10 ; thereafter, according to Figure 4.1.l(after NRC approval). The compilation of instrument failure rate data mcy include data obtained from other boiling water reactors for which the same design instrument operates in an environment similar to that of CNS.
2. Functional tests shall be performed before cach startup.with a required frequency not to exceed once per week.
3. Thir instrumentation is excepted from the functional test definition. The functional test will consist of applying simulated inputs. Local alarm lights representing upscale and downscale trips will be verified but no rod block w:.11 he produced at this time. The inoperative trip will be initiated ti produce a rod block (SRM and IRM inoperative also bypassed with the modo stitch in RUN). The functions that cannot be verified to produce a rod block directly will be verified during the operating cycle.
4. Simulated automatic actuation shall be performed once cach operating cycle.

Where possible, all logic system functional teuts will be performed using the test j acks.

5. Reactor low water level, high drywell pressure and high radiation main steam line tunnel are not included on Table 4.2.A since they are tested on Tabic 4.1.2.
6. The logic system functional tests shall include an actuation of time delay relays and timers necessary for proper functioning of the trip systems.
7. These units are tested as part of the Core Spray System tests.
8. The flow bias comparator will be tested by putting one flow unit in " Test" (producing a rod block) and adjusting the test input to obtain comparator l rod block. The flow bias upscale will be verified by observing a local upscale trip light during operation and verifying that it will produce a rod block during the operating cycle.
9. Performed during operating cycle. Portions of the logic is checked more f requently during functional tests of the functions that produce a rod block.
10. The detector will be inserted during each operating cycle and the proper amount of travel into the core verified.

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Attachment 2 Revised Technical Specifications for.

Definition of ' OPERABLE' Revised Page: 4 The current definition of "0PERABLE' in the Technical Specifications for Cooper Nuclear Station (CNS) states, " Operable means a system or component is capable of performing its intended function in its required manner." It does not provide sufficient detail to ensure the availability cf normal and

, emergency electrical power sources, controls, instrumentation, etc., that are required to support the object in performing its intended function under normal or unusual conditions. The need for clarification of the CNS Technical Specifications with regard to the definition of the term "0PERABLE" has been noted b) ;he NRC Resident Inspector as reported in NRC Inspection '

Report 84-04.

Nebraska Public Power District requests a revision to che Technical

. Specifications to incorporate a definition of the term "0PERABLE" that is in confornance with NUREG-0123, Revision 3 Standard Technical Specification 1.21

, which addresses the availability of auxiliary services and systems.

To increase the flexibility of applying this definition, yet remaining within its intent, the District requests a supplement to this definition called i " SAFETY REVIEWED OPERABLE". Under this definition a system, subsystem, etc..

I would be declared SAFETY REVIEWED OPERABLE and considered to have met the definition of OPERABLE despite the' existence of some apparent discrepancy if

, the Station Operation Review Committee (SORC) determines the effects of the discrepancy does not impair the function of the devise or system in question and does not conflict with any LCO.or other Technical Specification requirement. An example of a condition where this would apply is with a fault in the alarm circuitry of some reactor instrument yet the indication is unaffected. A special watch could be stationed to monitor the instrument and alert operators if the parameter reads certain values. Assuming other aspects of the instrument; i.e., inputs to the reactor protection system, etc., if any, are unaffected, the instrument, with the special watch stationed, could be declared SAFETY REVIEWED OPERABLE by SORC and meet the intent of the definition of OPERABLE.

j Evaluation of this Review with Respect to 10CFR50.92 j The proposed amendment incorporates a more restrictive definition of OPERABLE 1 to conform to GE Standard Technical Specifications and involves no significant j hazards considerations since it will not 1) involve a significant increase in

, the possibility or consequences of an accident previously evaluated, 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 j safety. The Commission has provided guidance concerning the application of the standards for making a no significant hazards consideration determination by providing certain examples (48FR14870). The exampics include "(ii) a change that constitutes an additional limitation, restriction, or control not  ;

j presently included in the Technical Specifications: for exampic, a more '

, stringent survelliance requirement." It is the District's belief the proposed l change is encompassed by the above example.

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K. Limiting Safety System Setting (LSSS) - The limiting safety system settings are set-tings on instrumentation which initiate the automatic protective action at a level such that the safety limits will not be exceeded. The region between the safety

, limit and these settings represent a margin with normal operation lying below these settings. The margin has been established so that with proper operation of the in-l strumentation the safety limits will never be exceeded.

L. Mode - The reactor mode is established by the mode selector switch. The modes in-clude refuel, run, shutdown and startup/ hot standby which are defined as follows:

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1. Refuel Mode - The reactor is in the refuel mode when the mode switch is in the 4

REFUEL position. When the mode switch is in the REFUEL position, the refueling interlocks are in service.

2. Run Mode - In this mode the reactor system pressure is at or above 825 psig and the reactor protection system is energized with APRM protection (excluding the 15% high flux trip) and RBM interlocks in service.

I 3. Shutdown Mode - The reactor is in the shutdown mode when the mode switch is in the SHUTDOWN position.

I 4. Startup/ Hot Standby Mode - In this mode the reactor protection scram trips initiated by the main steam line isolation valve closure are bypassed, the low pressure main steam line isolation valve closure trip is bypassed, the reactor protection system is energized with APRM (15% SCRAM) and IRM neutron monitoring system trips and control rod withdrawal interlocks in service.

I M. Operable - Operating - Inoperable

1. Operable - Operability - A system, subsystem, train, component or device shall d

be OPERABLE or have OPERABILITY when it is capable of performing its specified func tion (s) . Implicit in this definition shall be the assumption that all i

necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication ur other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).

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2. Safety Reviewed Operable - A system, subsystem, train, component or device may be declared " Safety Reviewed Operable" if it is capable of performing its spec-ified function even though an apparent discrepancy may exist in some associated instrumentation, control, power source, cooling or seal water, lubrication or j other auxiliary equipment. Such a determination is made by SORC review and approval and further requires that such a determination does not conflict with any limiting condition for operation nor any other technical specification re-quirement. Once a system, subsystem, train, component, or device is declared

, " Safety Reviewed Operable", it is considered OPERABLE per definition M.A.

3. Operating - Operating means a system, subsystem, train, component, or device is performing its intended function in its required manner.

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4. Inoperable - A system, subsystem, train, component, or device is inoperable if i

it is not capable of performing its intended function (s) in its required manner.

N. Deleted.

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0. Operating Cycle - Interval between the end of one refueling outage and the end of the
next subsequent refueling outage.

Attachtent 3 Revised Technical Specifications for NUREG-0737, Item II.K.3.18

" ADS Logic Modifications" Revised Pages: 59 76

Reference:

1) USNRC Letter. D. B. Vassallo to J. M. Pilant, NUREG-0737 Item II.K.3.18 " ADS LOGIC MODIFICATIONS", dated June 3, 1983.
2) NPPD Letter, J. M. Pilant to D. B. Vassallo. - same subj ect ,

dated July 22, 1983.

3) USNRC Letter, D. B. Vassallo to J. M. Pilant , same subj ec t ,

dated June 6, 1984.

Reference I required Nebraska Public Power District (NPPD) to chose one of two acceptable options for modifying ADS actuation logic to climinate the need for manual actuation for transient and accident events which do not directly produce a high drywell pressure signal. Reference 2 committed the District to modifying the ADS logic by eliminating the high drywell pressure permissive and adding a manual inhibit switch. Reference 3 stated the modification was found acceptabic by the staff and that the issue was closed.

Accordingly, NPPD requests a revision to the Technical Specifications to '

delete all references of 2 psig drywell pressure switches from Table 3.2.B (page 7) and Table 4.2.B (page 7) and to add surveillance requirements for the manual inhibit switches in Table 4.2.B (page 7).

Evaluation of this Revision with Respect to 10CFR50.92 The enclosed Technical Specification change is judged to involve no significant hazards based on the following:

1. Does the proposed license amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Evaluation:

Because the proposed change climinates the need for operator actuation of the ADS System for certain transient and accident events, it decreases the consequences of an accident previously evaluated and has no effect on its probability of occurring.

2. Does the proposed license amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Evaluation:

Because the proposed change does not introduce any new mode of operation, the possibility of an accident of a different type than analyzed in the Final Safety Analysis Report would not result from the change; therefore, the proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Evaluation:

Because the proposed amendment eliminates the need for operator actuation of ADS for certain events, thereby freeing the operator to monitor and evaluate the accident or transient and take actions to combat any-additional concerns, the proposed change does not reduce, but er.hances the margin of safety.

COOPER NUCLEAR STATION ,

TABLE 3.2.B (Page 7) .

Al' Tor.ATIC DEPRESSI'RI2ATION SYSTEM (ADS) CIRCUITRY REQUIREMENTS Minimum Number Action Required When Instrument of Operable Components Component Operability Irstrument I.D. No. Setting Limit Per Trip System (1) Is Not Assured Reactor Lev Eater NBI-LIS-83, A & B j +12.5" Indicated 1 B

'evel

- , Level NBI-LIS-72. A B.C & D > -145.5" Indicated 2 A Level ADS Timer MS-TDR-KS, A & B < 120 sec. I B Low-Low Set NBI-PS-51, A.B.C & D SI-A Open Low Valve 2 B 1015210 psig (Increasing) 51-B Close Low Valve 1 875!10 psig (Decreasing)

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51-C Open High Valve 1025210 psig (Increasing)

SI-D Close High Valve 875t10 psig (Decreasing)

COOPER NUCLEAR STATIC ^3 ,

TABLE 4.2.B (Page 7) .

ADS SYSTEM TEST & CALIBRATION FREQUENCIES Instrument Itc= Iten I.D. No. Functional Test Freq. Calibration Freq. Check Instrm,ca:s

1. ADS Inbtbit Switch MS-SW-S3A & B Once/ Month (1) N.A. None
2. Reacter Lew Water Level NLI-LIS-83. A & S Once/ Month (1) Once/3 Months Once/ Day NBI-LIS-72, A,B,C, & D Once/ Month (1) Once/3 Months Once/ Day
3. ADS. Ti:.er MS-TDR-K5 A & B once/ Month (1) Once/Oper. Cycle None
4. Lev-Lev Set NBI-PS-51, A.E.C. & D Gnce/ Month (1) Onec/Oper. Cycle None Lcgic (4)(6)
1. ADS Control Fever Monitor Once/6 Months N.A.
2. ADS Actuation Once/6 Months N.A.
3. Low-Low Set logic Once/6 Months N.A.

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Attachment 4 Revised Technical Specifications for Cooper Nuclear Station Organization Change Revised Page: 220 237 s

Nebraska Public Power District requests a revision to the Cooper Nuc1 car Station Technical Specifications to reflect changes in the organization for conduct of operation of the station shown in Figure 6.1.2. These changes include:

1. Title change from Special Projects Engineer to Senior Rad / Tech Advisor with broken line to Chemistry and itP Supervisor.
2. Titic change from Reactor Engineer Supervisor to Operations Engineer /

Supervisor.

3. Addition of Control Room Supervisor (SRO).

The title change from Reactor Engineer Supervisor to Operations Engineer Supervisor is also reficcted in 6.2.1.A.I.j on page 220.

4 Evaluation of this Revision with Respect to 10CFR50.92 The proposed amendment incorporates management organizational changes to improve the overall performance of Cooper Nuclear Station and involves no significant hazards consideration since it will not 11 involve a significant increase in the possibility or consequences of an accident previously

, evaluated, 2) create the possibility of a new or different kind of accident from any accident previounty evaluated, or 3) involve a significant reduction in a margin of safety. The Commission has provided guidance concerning the application of the standards for making a no signi'icant hazards consideration determination by providing certain exampics (48FR14870). The exampics include "

"(1) A purely administrative change to Technical Specificatiens: for example, a change to achieve consistency throughout the Technical Specifications, correction of an error, or a change in nomenclaturc." It is the District's beller the proposed change is encompassed by the above example.

6.2 REVIIM AND AUDIT

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6.2.1 The organization and duties of committees for the review and audit of station operation shall be as outlined below:

. A. Station Operations Review Committee (SORC) i

1. Membership:
a. Chairman: Division Manager of Nuclear Operations
b. Technical Staff Manager
c. Operations Manager
d. Technical Manager
c. Operations Supervisor j f. Maintenance Supervisor i g. Instrument and Control Supervisor
h. Chemistry and llealth Physics Supervisor l i. Plant Engineering Supervisor J. Operations Engineer Supervisor l ,

l k. Computer Applications Supervisor

1. Quality Assurance Manager - non-voting member.

Alternate members shall be appointed in writing by the Division 9 Manager of Nuclear Operations to serve on a temporary basis; however, 1

no more than two alternates shall serve on the Committee at any one i time.

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2. Meeting Frequency: Monthly, and as required on en11 of the Chairman.

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) 3. Quorum
Division Manager of Nuclear Operations or his designated alternate plus four other members including alternates.

j 4. Responsibilities:

a. Review all proposed normal, abnormal, maintenance and emergency

! operating prece ures specified in 6.3.1, 6.3.2, 6.3.3, and 6.3.4 l

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and proposed ch mges thereto: and any other proposed procedures or changes thereto determined by any member to ef fect nuclear safety.

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b. Review all proposed tests and experiments and their results, which '

l involve nuclear hazards not previously reviewed for conformance

' with technical specifications. Submit tests which may constitute an unreviewed safety question to the NPPD Safety Review and Audit j Board for review.

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c. Review proposed changes to Technical Specifications.

i d. Review proposed chanEus or modifications to station systems or j equipment as discussed in the SAR or which involves an unre-l viewed safety question ac defined in 10CFR50.59(c). Submit

] changes to equipment or systems having safety significance j

to the NPPD Safety Review and Audit Board for review.

l j e. Review station operation to detect potential nuclear safet)  :

hazarda. ,

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NUC L E AR OPER ATIONS DIVISION M AN AG E R STATION OPERATIONS FIRE PROTECTION REVIEW COMMITTEE -------

RM QUAllTY ASSURANCE TECH NIC A L

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M AN AGER - CNS MANAGER I I I I T R AINING TECHNICAL OPERATIONS ADMINIST R ATIV E SERVICES MANAGER M AN AGER MANAGER MANAGER I

OPE R AT IO N S S ENIO R I l TRAINING r-- RAD / TECH SECURITY ADMINIS TRATIVE M AT E RIAL SUPERVISOR I ADVISOR SUPPORT 4 SUPE 9 VISOR SU PERV ISO R SUPERVISOR J

r--

N  ! I I I l CHEM.& RR COMPUTER P L A NT OPER AT I ONS APPLICATIONS ENGINEER ENGINEER SUPER V ISO R SUP E R V ISO R SU PERVISOR SUPERVISOR I I l I/S ONE/ SHIFT 2/S TwO/ SHIFT M AINTEN ANCE I&C O P ER ATIONS 3/S THREE/ SHIFT SUPERVISOR SUPERVISOR SUPERVISOR (S R O )

RO-NRC REACTOR OPERATORS LICENSE SRO-NRC SENIOR REACTOR OPER ATORS l LIC EN S E u FUNCTIONAL POSITION ONLY I I SHIFT PHYSICALLY LOCATED IN CEERAL OFFICE ELECTRIC AL MAINT E NANCE MECH ANIC AL SUPERVISORS (SRO)

PLAN NER / I/S SU P ER VISO R SCHEDULER SUP ERVISOR l CONTROL ROOM SUPERV ISOR -

t/S

- Figure 6.1.2 I RRRD. Cooper Nuclear S totion 2/S [ UNIT OPER (RO) iN ER Organization Chart 3/S (STNiCE U ]

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