ML19317D230

Proposed Tech Specs 2.1,2.3,3.2,3.5 & 4.1 Required to Support Operation of Unit 1 at Full Rated Power During Cycle 5,including Core Protection Safety Limits & Protective Sys Mac Allowable Setpoint

ML19317D230
Person / Time
Site:	Oconee
Issue date:	06/26/1978
From:	DUKE POWER CO.
To:
Shared Package
ML19317D224	List: ML19317D223 ML19317D230
References
NUDOCS 7911190614
Download: ML19317D230 (20)

Text

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ATTACFSENT 1 Proposed Technical Specification Ravision Pages t

f l

i f

3911190 6/7

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. a can be related to DNB through the use of the BAW-2 correlation (1). The BAW-2 correlation has been developed to predict DNB and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core 1ccation to the actual heat flux, is indicative of the margin to DNB. The minimum value of the DNBR, during steady-state operation, normal operational transients, and. anticipated transients is limited to 1.30. A DNBR of 1.30 corresponds to a 95 percent probability at a 90 percent confidence level that DNB will not occur; this is considered a conservative margin to DNB for all operating conditions. The difference between the actual core outlet pressure and the indicated reactor coolant system pressure has been considered in determining the core protection safety limits. The difference in these two pressures is nominally 45 psi; however, only a 30 psi drop was assumed in reducing the pressure trip setpoints to correspond to the elevated location where the pressure is actually measured.

The curve presented in Figure 2.1-1A represents the conditions at which a minimum DNBR of 1.30 is predicted for the maximum possible thermal power (112 percent) when four reactor coolant pumps are operating (minimum reactor coolant flow is 106.5 percent of 131.3 x 10' lbs/hr.). This curve is based on the combination of nuclear power peaking factors, with potential effects of fuel densification and rod bowing, which result in a more conservative DNBR than any other shape that exists during normal operation.

The curves of Figure 2.1-2A are based on the more restrictive of two thermal limits and include the effects of potential fuel densification and rod bowing:

1. The 1.30 DNBR limit produced by the combination of the radial peak, axial peak and position of the axial peak that yields no less than a 1.30 DNBR.

2. The combination of radial and axial peak that causes central fuel melting-

! at the hot spot. The limit is 20.15 kw/ft for Unit 1.

Power peaking is not a directly observable quantity and therefore itnits have been established on the bases of the reactor power imbalance produced by the power peaking.

The specified flow rates for Curves 1, 2, and 3 of Figure 2.1-2A correspond to the expected minimum flow rates with four pumps, three pumps, and one pump in each loop, respectively.

The curve of Figure 2.1-1A is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in Figure 2.1-3A.

The maximum thermal power for three-pump operation is 85.3 percent due to a l power level trip produced by the flux-flow ratio 74.7 percent flow x 1.055 =

l 78.8 percent power plus the maximum calibration and tastrument error. The maximum thermal power for other coolant pump conditions are produced in a similar manner.

2.1-2

) C Thermal Power Level, 7,

(-28,i12) (32,i12)

~

I ACCEPTABLE

/(.yo,gg) - 100 4 PUMP \,(50,98)

OPERAT10N (32,85.3) 2

(-28,85.3) --

80 .

3&4 PUMP (50,71.3)

(-40,71.3) OPERATION

. 60 (32,58.2) ,,

3 ,

(-28,58.2)

ACCEPTABLE i (-40,44.2) 2,3&4 PUMP ( ' '

_. 40 OPERA, T ION 20 l

I I I f f f i 60 -40 -20 0 20 40 60 80 Reactor Powei imbalance, 7 l

CURVE RC FLOW (GPM) l 374,880 2 280,035 3 183,690 CORE PROTECTION I SAFETY LIMITS l

. UNIT 1 a

g OCONEE NUCLEAR STATION Figure 2.1-2A n e ,

g 2440 Act!PTA81.E '

230 ~

a OPEAATION f ,

E l $

\ 2000 )

1 2 5

=

O 1800

/ #

-l l 1600 580 580 $60 620 640 Reactor Csolant Outist Tasceraturs-F PUMPS TYPE OF CURVE REACTOR COOLANT FLOW (GPM) POWER OPERATING LIMIT 1 374880 (100%)* 112% 4 (DNBR) 1 2 280035 (74.7%) 85.3% 3 (DNBR) 3 183650 (49 0%) 58.2% 2 (QUALITY) l *106.5% of first core design flow CORE PRCTECTICN SAFETY LIMITS UNIT 1 laant.m t CCONEE NUCLEAR STATION Figure 2.1-3A 2.1 - I.0

)

Thermal Power, %

(-l8,105.5) p (20,105.5)

Mg = 0.558 4 PUMP 3

- 100 \

l (30,98)

(-40,93) OPERATION I e

~

x (-l8,78.8) __ 80 (20.78.8) . =

b l\ (30,71.3) 2 (-40,66.3) 3&'4 PUMP  %

E OPERATION E l' --

60 l w u,

g y (-18,51.69) (20,51.69) y 5 5 g

g - (-40,39.19) l l

-- 40 Nl (30,44.19) 8 g

1 J l l o "i $ = I o - -

20 o l o '

E l n g,= l: si  :

E E m o ,l E E l E, .

-60 -40 -20 0 20 40 60 Reactor ' Power Imbalance, %

l PROTECTIVE SYSTEM MAXIMUM ALLCWASLE SETP0lNTS

. UNIT I 8

OCONEE NUCLEAR STATION

. Figure 2.3-2A 2.3-8

--- ------....---,.,-,,-p , , - , _ , , , , , , , , , _ , _ , , , , . , , _ , ,

r, s

Bases The high pressure injection system and chemical addition system provide control of the reactor coolant system boron concentration. (1) This is normally accom-plished by using any of the three high pressure injection pumps in series with a boric acid pump associated with either the boric acid mix tank or the concen-trated boric acid storage tank. An alternate method of boration will be the use of the high pressure injection pumps taking suction directly from the borated water storage tank. (2)

The quantity of boric acid in storage in the concentrated boric acid storage tank or the borated water storage tank is sufficient to borate the reactor coolant system to a 1%Ak/k suberitical margin at cold conditions (70 F) with the maximum worth stuck rod and no credit for xenon at the worst time in core life. The current cycles for each unit, Oconee 1 Cycle 5, Oconee 2, Cycle 3, and Oconee 3, l Cycle 4 were analyzed with the most limiting case selected as the basis for all three units. Since only the present cycles were analyzed, the specifications will be re-evaluated with each reload. A minimum of 980 fta of 8,700 ppm boric acid in the concentrated boric acid storage tank, or a minimum of 350,000 gallons of 1800 ppm boric acid in the borated water storage tank (3) will satisfy the re-quirements. The volume requirements include a 10% margin and in addition allow for a deviation of 10 EFPD in the cycle length. The specification assures that two supplies are available whenever the reactor is critical so that a single failure will not prevent boration to a cold condition. The required amount of boric acid can be added in several ways. Using only one 10 gpm boric acid pump taking suction from the concentrated boric acid storage tank would require ap-proximately 12.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> to inject the required boron. An alternate method of addition is to inject boric acid from the borated water storage tank using the makeup pumps. The required boric acid can be injected in less than six her,rs using only one of the makeup pumps.

The concentration of boron in the concentrated boric acid sto: age tank may be higher than the concentration which would crystallize at ambient conditions.

For this reason and to assure a flow of boric acid is available when needed, these tanks and their associated piping will be kept at least 10 F above the crystallization temperature for the concentration present. The boric acid con-centration of 8,700 ppm in the concentrated boric acid storage tank corresponds to a crystallization temperature of 77 F and therefore a temperature requirement of 87 F. Once in the high pressure injection system, the concentrate is suffi-ciently well mixed and diluted so that normal system temperatures assure boric acid solubility.

REFERENCES (1) FSAR, Section 9.1; 9.2 (2) FSAR, Figure 6,2 (3) Technical Specification 3.3 3.2-2

3 m

. J _)

3.5.2.5 Centrol Rod Pcsitions

a. Technical Specification 3.1.3.5 does not prohibit the exercising of individual safety rods as required by Table 4.1-2 or apply to in-operable safety rod limits in Tashalcal Specification 3.5.2.2.

b. Except for physics test, operating rod group overlap shall be 25% + 5% between two sequential groups. If this limit is exceeded, corrective measures shall be taken immediately to achieve an ac-ceptable overlap. Acceptable overlap shall be attained within two hours or the reactor shall be placed in a hot shutdown condition within an additional 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c. Position limits are specified for regulating and axial power shaping control rods. Except for physics tests or exercising control rods, the regulating control rod insertion / withdrawal limits are specified on figures 3.5.2-1A1 and 3.5.2-1A2 (Unit 1); 3.5.2-1B1, 3.5.2-1B2 and 3.5.2-1B3 (Unit 2); 3.5.2-1C1, 3.5.2-1C2 and 3.5.2-1C3 (Unit 3) for four pump operation, and on figures 3.5.2-2A1 and 3.5.2-2A2 (Unit 1); 3.5.3-231, 3.5.2-2B2 and 3.5.2-2B3 (Unit 2); 3.5.2-2C1, 3.5.2-2C2 and 3.5.2-2C3 (Unit 3) for two or three pump operation.

Also, excepting physics tests or exercising control rods, the axial power shaping control rod insertion /vithdrawal limits are specified on figures 3.5.2-4A1, and 3.5.2-4A2 (Unit 1); 3.5.2-4B1, 3.5.2-4B2, and 3.5.2-4B3 (Unit 2); 3.5.2-4C1, 3.5.2-4C2, and 3.5.2-4C3 (Unit 3).

If the control rod pcsition limits are exceeded, corrective measures shall be taken immediately to achieve an acceptable control rod posi-tion. An acceptable control rod position shall then be attained with-in two hours. The minimum shutdown margin required by Specification i

3.5.2.1 shall be maintained at all times.

3.5.2.6 Xenon Reactivity

a. Except for physics tests, rea'ctor power in Unit I shall not be in-creased above the power-level-cutoff shown in Figures 3.5.2-1A1, and 3.5.2-1A2 unless one of the following conditions is satisfied:

1. Xenon reactivity did not deviate more than 10 percent from the i equilibrium value for operation at steady state power.

i

2. Xenon reactivity deviated more than 10 percent but is now with-in 10 percent of the equilibrium value for operation at steady state rated power and has passed its final maximum or minimum peak during its approach to its equilibrium value for operation at the power level cutoff.

3. Except for xenon free startup (when 2. applies), the reactor has l operated within a range of 87 to 92 percent of rated thermal power for a period exceeding 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in the soluble poison control mode.

i

b. Except for physics tests, reactor power in Units 2 and 3 shall not be increased above the power level cutoff shown in Figures 3.5.2-1B1, l

3.5.2-1B2, and 3.5.2-1B3 (Unit 2); 3.5.2-1C1, 3.5.2-1C2, 3.5.2-1C3 (Unit 3); unless the following requiremects are met:

3.5-9

1. The xenon reactivity shall be within 10 percent of the value for operation at steady-state rated power.

2. The xenon reactivity worth has passed its final maximum or minimum peak during its approach to its equilibrium valve for operation at the power level cutoff.

3.5.2.7 Reactor power imbalance shall be monitored on a frequency not to exceed two hours during power operation above 40 percent rated power. Except for physics tests, imbalance shall be maintained within the envelope defined by Figures 3.5.2-3A1, 3.5.2-3A2, 3.5.2-3B1, 3.5.2-3B2, 3.5.2-3B3, 3.5.2-3C1, 3.5.2-3C2, and 3.5.2-3C3. If the imbalance is not within the envelope defined by these figures, corrective measures shall be taken to achieve an acceptable imbalance. If an acceptable imbalance is not achieved within two hours, reactor power shall be reduced until imbalance limits are met.

3.5.2.8 The control rod drive patch panels shall be locked at all times with limited access to be authorized by the manager or his designated alternate.

3.5-10

Bases .

The power-imbalance envelope defined in Figures 3.5.2-3A1, 3.5.2-3A2, 3.5.2-3B1, l 3.5.2-3B2, 3.5.2-3B3, 3.5.2-3C1, 3.5.2-3C2 and 3.5.2-3C3 is based on LOCA analyses which have defined the maximum linear heat rate (see Figure 3.5.2-5) such that the maximum clad temperature will not exceed the Final Acceptance Criteria.

Corrective measures will be taken immediately should the indicated quadrant tilt, rod position, or imbalance be outside their specified boundary. Opera-tion in a situation that would cause the Final Acceptance Criteria to be ap-proached should a LOCA occur is highly improbable because all of the power distribution parameters (quadrant tilt, rod position, and imbalance) must be at their limits while simultaneously all other engineering and uncertainty factors are also at their limits.** Conservatism is introduced by application of:

a. Nuclear uncertainty factors

b. Thermal calibration

c. Fuel densification power spike factors (Units 1 and 2 only)

d. Hot rod manufacturing tolerance factors

e. Fuel rod bowing power spike factors The 25% 1 5% overlap between successive control rod groups is allowed since the worth of a rod is lower at the upper and lower part of the stroke. Control rods are arranged in groups or banks defined as follows:

Group Function 1 Safety 2 Safety 3 Safety 4 Safety 5 Regulating 6 Regulating 7 Xenon transient override 8 APSR (axial power shaping bank)

The rod position limits are based on the most limiting of the following three

criteria

ECCS power peaking, shutdown margin, and potential ejected rod worth.

l Therefore, compliauce with the ECCS power peaking criterion is ensured by the rod position limits. The minimum available rod worth, consistent with the rod position limits, provides for achieving hot shutdown by reactor trip at any I time, assuming the highest worth control rod that is withdrawn remains in the

( full out position (1). The rod position limits also ensure that inserted rod groups will have been shown to be safe by the safety analysis (2,3,4,5) of hy-

! pothetical rod ejection accident. A maximum single inserted control rod worth

! of 1.0%Ak/k is allowed by the rod position limits at hot zero power. A single inserted control rod worth of 1.0%Ak/k at beginning-of-life, hot zero power would result in a lower transient peak thermal power and, therefore, less severe environmental consequences than a 0.65%ak/k ejected rod worth at rated power.

• • Actual operating limits depend on whether or not incore or excore detectors are used and their respective instrument calibration errors. The method used

to define the operating limits is defined in plant operating procedures.

3.5-11 .

i l

I

-..e---- ---

.m s

, Control rod groups are withdrawn in sequence beginning with Group 1. Groups 5,6, and 7 are ' overlapped 25 percent. The normal position at power is for Groups 6 and 7 to be partially inserted.

The quadrant power tilt limits set forth in Specification 3.5.2.4 have been es-tablished to prevent the linear heat rate peaking increase associated with a

, positive quadrant power tilt during normal power operation from exceeding I

7.50% for Unit 1. The limits shown in Specification 3.5.2.4 l 1

5.10% for Unit 2 7.50% for Unit 3 are measurement system independent. The actual operating limits, with the appro-

, priate allowance for observability and instrumentation errors, for each measure-i ment system are defined in the station operating procedures.

The quadrant tilt and axial imbalance monitoring in Specification 3.5.2.4 and 3.5.2.7, respectively, normally will be performed in the process computer. The l two-hour frequency for monitoring these quantities will provide adequate sur-veillance when the computer is out of service.

j Allowance is provided for withdrawal limits and reactor power imbalance limits to be exceeded for a period of two hours without specification violation. Accept-able rod positions and imbalance must be achieved within the two-hour time period or appropriate action such as a reduction of power taken.

Operating restrictions are included in Technical Specification 3.5.2.6 to prevent l l excessive power peaking by transient xenon. The xenon reactivity must be beyond i its final maximum or minimum peak and approaching its equilibrium value at the power level cutoff.

REFERENCES I FSAR, Section 3.2.2.1.2 FSAR, Section 14.2.2.2 3

! FSAR, SUPPLEMENT 9

'B&W FUEL DENSIFICATION REPORT BAW-1409 (UNIT 1)

BAW-1396 (UNIT 2)

BAW-1400 (UNIT 3)

~

5

0conee 1, Cycle 4 - Reload Report - BAW-1447, March 1977, Section 7.11.

l 3.5-11a

) em J )

TABLE 3.5-1 Quadrant Power Tilt Limits 1 Steady State Transient Maximum i

Limit Limit Limit i Unit 1 5.00 9.44 20.0 Unit 2 3.41 9.44 20.0 Unit 3 5.00 9.44 20.0 1

f 1

i i

i I

i.

l l

.l h

I I

3.5-11b

d'

] )

(125,102) (274.1,102) 100 . O OPERATICM NOT ALLOWED (274.1,92) __

- power 80 E '.

(248.2,80)

RESTRil:TD 3MUTDOWN MARGIN LIMIT REGION 60 .

(214.7,60) 1

• (70,50)

A

, 50 -

I PUNf33 ISLE 0 .RATIMO 20 -(18.15) REGICM (0,12.5 O, ' ' ' , . , '

L i , , ,

u 50 100 150 200 250 3C0 Rod indez, 5 WO O 25 50 75 100 , 0 25 50 75 100 i , . . t , . , , ,

Group 5 Group 7 0 25 50 75 100 e r . t t Group 6 ROD POSITION LIMITS FOR FOUR PUMP OPERATION FROM 0 TO 100 + 10 EFP0 -

UNIT 1 asu .m* CCONEE NUCLEAR STATION Figure 3.5.2-1A1 3.5-12 I

(241.102) (274.1.102) 100 - O OPE m :0" (272.1,s2) e _ _ _ _

NOT ALLOWED PCwER 248.2.50) tg.iEL 80 -

CUTOFF

= 92". FP j 60 -

+

(214.7.50) y (179.50) f PERMi$315Li 40 -

==.

CPfRATlfqG j

a.

SHUTDCWM WARGIN LIMIT -

REGION 20 (103,15) ,

,(0.6.1) 0_ ,(0.0) , , , , , , , , , ,

0 50 100  !!O 200 250 200 Rod index. #. Withdrawn 0 25 50 75 100 - 25 i . . . ,0 . 5,0 75, 10,0 Grous 5 Group 7 0 25 50 75 ICO

, , , , i Grsup 5 ROD POSITION LIMITS FOR FOUR PUMP OPERATION AFTER 100 + 10 E.:PO -

UNIT 1 --

o OCONEE NUCLEAR STATION Figure 3.5 2-1A2 3.5-13

m., .3

) ]

(125.102) (198,102 (2'48.2.102)

'22 .

OPERATION NOT ALLChED g@

50 -

g

//

Df I 4 (211. 7,75) ,

(120,61) t 5HtJTDOWN MARGIN j ~

LIMIT s@ g@

(70,50) g@

j PERNISSISLE 40 -

, OPERATING REGION (18.15)

• O 20 h.15 %

a (0,0) , , , , , r o ,

t , ,

0 50 100 150 200 250 300 Rod Index. '. Withdrawn 0 25 5,0 75 0 100 1,00 , 2,5 5,0 7,5 Group 5 Group 7 0 25 50 75 100 i . . . .

Group 5 I

(

ROD POSITION 6 'MITS FOR TWO AND THREE PUMP OPERATION FROM 0 70 100 1 10 EFPD UNIT 1 OCONEE NUCLEAR STATION Figure 3.5 2-2Al 3.5-13

, n.

J .)

e, EuTio, (2ii.iO2) (2s.2.ic2) 100 -

NOT ALLCwtll RESTRICTED FOR 3 PtMP 80 -

214.7,76) 2

~

8 60 -

(179,50) PERMIS3 ISLE OPERATING E SMUTDOWN MARolN LlMIT -

, RE31CN

, 30 -

t 5

20 -

(103.151 (0,s.1) 0 t . , , , . . , , , ,

Rod .Index, hit.Sdrawn 50 3C0 25 5 " 'a 2,5 n ico

P s.O Grcup 5 Group 7 0 50 75 100 t 2,5 t e Group 6 l

l l

ROD POSITION LIMITS l FOR T'40 AND THREE PUMP OPERATION AFTER 100 + 10 EFPD

~

UNIT 1 l OCONEE NUCLEAR STATION Figure 3.5.2-2A2 l

3 5-13a

Pow r, 7. of 2568 HWt

~

RESTRICTED REGION

(-26.9,iO2'r .

- 100

(-28.2,92) "

90

(-29.2,80) <' -- 80 "

(15.9,80) ,

- - 70 PERMI SSI BLE -- 60 OP ERATING REGION

-- 50

- - 40

-- 30

- - 20 10 i r r r r _ 1 i , e r

, -50 -40 -30 -20 -10 0 +10 +20 +30 +40 +50 l Axial Power Imbalance, f.

l l

1 I

OPERATIONAL POWER IMSALANCE ENVELOPE FOR OPERATION FROM 0 to 100 - 10 EF?D UNIT 1

'o

, OCONEE NUCLEAR STATION Figure 3.5.2-3A1 3.5-21

y Powar, 7. of 2568 HWt

~

--110 RESTRICTED REGION

(-28.0,102) ,

(16.4,102)

(-28.5,92) 0

- 90

(-29.2,80)o - . 80 " (l6.4,80)

PERMISSIBLE --

70 OPERATING REGION --

60

- - 50 40 30

-. 20

• . . 10

\ .

-50 -40 -30 -20 -10 0 10 20 30 40 50 Axial Power imbalance, 7 l

l OPERATIONAL POWER IMBALANCE ENVELOPE I FOR OPERATICN AFTER 100 + 10 EFFD -

l UNIT 1 o OCONEE NUCLEAR STATION Figure 3.5.2-3A2 3.5-21a

b ...

(16.3,102) (45.0,102)

RESTRICTED (l4.!,92) - (45.0,92)

'80 *'

(0.0,80)

(100,60) 1 l

3 60 -

E N

PERMISSIBLE

, d 0 -

OPERATING ,

l

,- REGION e

i 3

n. '

20 0

0 20 40 60 80 100 Sank 8 Position, ". Withdrawn 1

1 APSR POSITION LIMITS FOR OPERATICN FRCM 0 to 100 1 10 EFPD UNIT 1

) OCONEE NUCLEAR STATION Figure 3.5.2-hat i

1 3.5-23c

) J (9.3,102) -

(45.0,102) o R'STRICTED 100 -

dEGION (7.6,92) , (45.0,92) 80 (0.0,80) (57.9,80)

~

100,60) 5 60 -

PERl41SSIBLE h

OPERATING

• REGION ,

d 40 -

a.

20 -

I .

'-~----' ' -

r t 0

40 60 80 100 t 0 20 t

Bank 8 Position, f. Withdrawn l -

APSR POSITION LIMITS FOR OPERATION AFTER 100 + 10 EFPO -

-- Uni: 1 J OCONEE NUCLEAR STATION Figure 3.5.2-LA2 3.5-23d

4.1 -

OPERATIONAL SAFETY REVIEW Applicability Applies to items directly related to safety limits and limiting conditions for operation.

Obj ective To specify the frequency and type of surveillance to be applied to unit equip-ment and conditions.

Specification 4.1.1 The frequency and type of surveillance required for Reactor Protective System and Engineered Safety Feature Protective System instrumentation shall be as stated in Table 4.1-1.

4.1.2 Equipment and sampling test shall be performed as detailed in Tables 4.1-2 and 4.1-3. '

4.1.3 Using the Incore Instrumentation System, a power map shall be made to verify expected power distribution at periodic intervals not to exceed ten effective full power , days.

Bases Failures such as blown instrument fuses, detective indicators, and faulted amplifiers which result in " upscale" or "downscale" indication can be easily recognized by simple observation of the functioning of an instrument or sys-tem. Furthermore, such failures are, in many cases, revealed by alarm or annunciator action. Comparison of outp,ut and/or state of independent channels measuring the same variable supplements this type of built-in surveillance.

Based on experience in operation of both conventional and nuclear systems, when the unit is in operation, the minimum checking frequency stated is deemed adequate for reactor system instrumentation.

Calibration is performed to assure the presentation and acquisition of accurate information. The nuclear flux (power range) channels amplifiers are calibrated (during steady-state operating conditions) when indicated neutron power exceeds core thermal power by more than two percent. During non-steady-state opera-tion, the nuclear flux channels amplifiers are calibrated daily to compensate for instrumentation drift and changing rod patterns and core physics parameters.

Channels subj ect only to "drif t" errors induced within the instrumentation it-self can tolerate longer intervals between calibrations. Process system instru-mentation errors induced by drif t can be expected to remain within acceptable tolerances if recalibration is performed at the intervals specified.

Substantial calibration shifts within a channel (essentially a channel failure) l are revealed during routine checking and testing procedures. Thus, the minimum

' calibration frequencies set forth are considered acceptable.

i 4.1-1

• w- , - - - . - -