ML20085H440

Proposed Tech Spec 3.1.B.1 Re Min Number of APRM Channel Inputs Required to Permit Accurate Average Core Power Monitoring

ML20085H440
Person / Time
Site:	Oyster Creek
Issue date:	10/09/1991
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20085H432	List: ML20085H428 ML20085H434 ML20085H440 ML20085H450
References
NUDOCS 9110280277
Download: ML20085H440 (5)

Text

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

l 3.1 PROTECTIVE INSTRUMFNTATION I

hpolicability: Applies to the operating status of plant instrumentation which performs a protective function.

Obiective: To assure the operability of protective instrumentation. j i

Snecifications: h. The following operating requirements for plant protective instrumentation are given in Table 3.1.1:

1. The reactor mode in which a specified function must be operable including allowab1w bypass conditions.

2. The mintmum number of operable instrument channels por operable trip system.

3. The trip settings which initiate automatic protective action.

4. The ,cion required when the limiting conditions for ,

operation are not satisfied. I B. 1. Failure of four chambers assigned to any one APRM shall make the APRM inoperable. l

2. Failure of two chambers from one radial core location in any one APRM shall make thst APRM inoperable.

3. Except during the performanco of Technical specification required LPRM/APRM surveillance, reactor power shall be reduced below the 80% rod line or the corresponding RPS trip siystem shall be placed in the tripped condition, whenever all three of the following conditions exist

1. Reactor Power is greater than 35%

-and-

2. More than one LPRM detector is bypassed or failed in the A level or the B level assigned to a single APRM channel

-and-

3. the diagonally opposite quadrant contains a single APRM channel with more than one bypassed or failed LPRM detector on the same. axial level as the bypassed or f ailed detectors specified ira (2) above.

OYSTER CREEK 3.1-1 Amendment No.: 64 l

9110280277 911009

-PDR ADOCK 05000219 P PDR

-l l

l 1

T[  : C. 1. JAny'two (2) LPRM assemblies which'aro input to the-APRM system and are~ separated in distance by less than ,

three (3) times the control rod pitch may 'not contain a combination of more than three (3) inoperable detectors (i.e., APRM channel failed or bypassed, or LPRM detectors failed or bypassed) out of the four.(4) detectors located in either the A and B, or the C and, -]

D levels. l 2 A Travelling In-core Probe (TIP) chamber may be insed i as an APRM input to meet the criteria of 3.1.D and 3.1.C,1,_ provided the TIP is positioned in close proximity to one of the failed LPRM's. If the ,

criteria of 3.1.B.2 or 3.1.c.1.cannot be met, power operation may continue at up to rated power level

~

provided a control rod withdrawal block is operati.ng-or at power levels less than 61% of rated power until

~

the TIP can be connected, positioned and  ;

satisf actorily test.ed, as long as Specification 3.1.B.1 and Table 3.1.1 are satisfied.

.Banen: The plant protection system automatically initiates protective

. functions to prevent exceeding established limits. In addition, other

-protective inctrumentation'is provided to initiate action which-mitigates:the consequences of accidents or terminates operator control. 'Tnis specification provides the limiting conditions for operation necessary to-preserve the effectiveness of these instrument systems, t

OYSTER CREEK 3.1-1(a) Amendment No.: 64 ,

. a .. _ _ . . .. _ . . . _ _ _ . _ . . _ _ _ ,

r b_

Ranges 8 and. higher.--It is not required at this power level since good.

Indication-exists in the Intermediate Range and the SRM will be reading 5

-approximately=5-x 10 CPS when using IRM Ranges 8 and higher.

-The IRM downscale rod block:in conjunction with the chamber full-in position and range switch setting, provides a rod block to assure that the IRM is in its most sensitive condition before startup, -If the two latter conditions are satisfied, control rod withdrawal may commence even if_the IRM is not reading at-least-5%; However, after a substantial neutron flux is obtained, the rod block setting prevents the chamber from being withdrawn to an insensitive area

-of the core.

The APRM downscale setting of 2 2/150 full scale is provided in the run mode to prevent control rod withdrawal without adequate neutron monitoring.

High flow lin the main steamline is set at 120% of rated flow. At this retting the isolation valves close and in the event of a steam line' break limit the

-loss of--. inventory so:that-fuel clad perforation does not occur.- The 120% flow

-would correspond.to the thermal power so this would either indicate a line break or too'high a power.

Temperature sensors aca provided in the steam line tunnel to provide for closure of- the main steam.*ne isolation valves should a break or leak occur in this area of the= plant. The trip is' set at 50*F above ambient temperature at rated power. This setting will cause isolation to occur for main steamline

~

breaks which result in a flow of a-few pounds per minute or greater. Isolation occurs soon.enough to meet the criterion of no clad perforation.

The-low-low-low water level trip point is set at 4'8" above the top of the active. fuel-and will prevent spurious operation of the automatic relief system. The trip point established will initiate the automatic

~

depressuritation system 1n time to provide adequate core cooling. _

Specification 3.1.D.1 defines the minimum number of APRM channel inputs required to permit accurate average. core power monitoring.- Specification 3.1.B.3 defines APRM channel inputs operability requirements in order:to ensure a sufficient-APRM response to regional power oscillations. Specifications l 3.1.B.2 and 3.1.C.1 further define the distribution of the operable chambers to l provide monitoring of local power. changes that might be caused by a single rod withdrawal. Any nearby, operable LPRM chamber can provide the required input

~for-average core monitoring. A Travelling Incore Probe or Probes can be used y temporarily to provide APRM input (s) until LPRM replacement is-possible. Since t

lAPRM rod block protection is not required below 61% of rated power, (1) as

-discussed in-Section 2.3, I lmitina -Sa f ety System Settings, operation may continue below 61% as long as Specification 3.1.B.1 and the requirements of Table'3.1.1 are met. In order-to maintain reliability of core monitoring in

-that quadrant where an APRM isfinoperable, it is permitted to remove the operable APRM from service for calibration and/or test provided that the same core protection is maintained _by. alternate means.

In the rare ovent that Travelling In-core Probes (TIPS) are used to meet the

, requirements 3.1.B or 3.1 C, the licensee may perform an analysis of substitute l' LPRM inputs- to the: APRM system using spare (non-APRM input) LPRM detectors and' change the APRM system as permitted by 10 CFR 50.59.

Change: 6

-OYSTER CREEK 3.1-6 Amendment No.: 9, 15, 112 01/4/80 1

Eb _-

y LC. (Minimum. Critical-Power Ratio (MCPR). I

~During steady state power operation the MINIMUM CRITICAL POWER RATIO ,

(MCPR) shall be' equal to or great er than the MCPR limit as specified -l in the COLR.

5e MCPR' limit for each cycle as identified in the COLR shall be greater than or. equal to 1.47.

When APRM status changes _ duo to instrument faAlure (APRM or LPRM1 input s failure), the MCPR requirement for the degraded condition shMll be met .

z within a time interval ~of eight (8) hours, provided that the con' trol- l rod block is placed-in operation during this interval. >

For. core flows other than rated, the nominal value for MCPR shall be increased by a factor of k g, where kg is as shown in the COLR. ,

If-at any time during power operation it is determined by normal .

surveillance that the limiting value for MCPR is being exceeded for reasons other than instrument failure, action shall be initiated to restore operation to within the prescribed-limits. If the steady state- MCPR is not returned to within the prescribed limits within two (2) hours,': action shall be initiated to bring the reactor to.the cold '

shutdown condition within 36 hourn. During this period, surveillance and corresponding action shall continue until reactor operation is within the prescribed limit at which time power operation nay be continued.

Basesi The Specification for average planar LHGR assures that the po4k

= cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the 2200*F limit specified in

- 10 CFR 50.46. The analytical methods and assumptions used'in evaluating the fuel design limits are presented in FSAR Chapter 4. ,

LOCA analyses are performed for_each fuel' design at selected exposure L -points.to determined APLHGR limits that meet-the PCT and maximum L oxidation limits of 10 CFR 50.46. The analysis is performed using GE

!- calculational models which are consistent with the requirements of 10 CFRv50, Appendix K.

- The PCT following.a postulated LOCA is primarily a function of the L .-average heat generation rate of all the rods of a fuel assembly at-any

- axia1L-location and-is not strongly influenced by the rod to rod power distribution within an~ assembly. Since expected location variations-in powor-distribution within a. fuel assembly affect the calculated peak clad temperature by less than 1 20*F relative to the peak temperature for a typical fuel design,' the limit on the average linear heat generation rate is sufficient to assure that calculated l temperatures are-below the limits specified in 10 CFR 50.46.

L _ .

l -

OYSTER CREEK 3.10-2 Amendment No.: 48, 75, l

lil, 129, 147

The maximum averega planar:LHGR limits for the various fuel types currently being used are provided~in the COLRt The l l MAPLHGR limite.for both five-loop and four-loop operation with the idle loop unisolated'are shown. -Four-loop operation with the idle loop isolated (suction, discharge and discharge bypass valves closed) requires that a RAPLHCR multiplier of O.98 be applied to all fuel types.

Additional requirements for-isolated-loop operation are given in Specification 3.3.F.2.

Fuel design evaluations are performed to demonstrate that the cladding is plastic strain and other fuel design limits are not' exceeded during anticipated operational occurrences for operation with LHGRs up to the operating limit LHGR.-

~

The analytical methods and assumptions used in evaluating the anticipated operational occurrences to establich the operating limit MCPR are presented in the FSAR, Chapters 4, H

6 and 15 and in Technical-Specification 6.9.1.f. To assure that: the ' Safety Limit MCPR is not exceeded during any ,

moderate frequency transient event, limiting transients -

have been analyzed to' determine the largest reduction in Critical Power Ratio (CPR).- The types of transients-  ;

evaluated are pressurization, positive reactivity insertion and coolant temperature decrease. The operational MCPR-limit is selected to provide margin to accommodate transients and uncertainties in monitoring the core operating state, manufacturing, and in the critical power correlation itself. =This limit is derived by addition of the- CPR for the most limiting transient to the safety limit MCPR designated in Specification 2.1.

t A lower bound of 1.47 has been established for the operating limit MCPR=value to provide sufficient margin to ,

the MCPR safety limit in the event of reactor thermal-hydraulic. instability. The 1.47 limit will be considered against the minimum operating CPR limit ~ based on reload transient and accident analysis,. The higher of core stability or. reactor transient and accident determined.MCPR-will.be used to determino the cycle operating limit.

The-APRM response is used'to predict when the rod block occurs in the analysis of the rod withdrawal error transient. The transient rod position at the rod _ block and corresponding MCPR can be determined. The MCPR has been.

evaluated for different APRM responses which would reault -

from changes in the.APRM etatus as a. consequence of bypassed APRM channel and/or; failed / bypassed LPRM inputs. .

, The steady state MCPR required to protect the minimum transient CPR for the worst case APRM status condition (APRM Status 1) is determined in the rod withdrawal error h transient analysis. .The steady state MCPR values for APRM status conditions 1, 2, and 3 will be evaluated each cycle. For those cycles where the rod withdrawal error

! . transient is not the most severe transient the MCPR Value for APRM status conditions 1, 2, and 3 will be the same and

-be equal to the limiting transient MCPR-value.

l

- OYSTER CREEK- 3.10-3 Amendment No. 75, 129, 147 i'

I E u