ML19347F653

Proposed Tech Specs 2.2.2,6.7,6.8,3.2.7,4.2.6,3.10.2,4.10.2 & 4.3.1 for Cycle 2 Reload

ML19347F653
Person / Time
Site:	Arkansas Nuclear
Issue date:	05/19/1981
From:	ARKANSAS POWER & LIGHT CO.
To:
Shared Package
ML19347F652	List: ML19347F651 ML19347F653
References
NUDOCS 8105220350
Download: ML19347F653 (15)

Text

s 4

llh SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS (CONTINUED)

CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS 2.2.2 Core Protection Calculator- Addressable Constants are defined in Table 2.2-2. Type I Addressable constants are expected to change frequently during plant operation. Type II Addressable constant values are determined (or confirmed) during PHYSICS TESTS following each fuel loading and are not expected to change during plant operation.

Changes to Type I Addressable constants outside the Allowable Value range require Plant Safety Committee review prior to implementation.

Changes to Type II Addressable constants made other than as a result of post fuel loading PHYSICS TESTS shall require Plant Safety Committee review prior to implementation unless the changes are required for Technical Specification Compliance.

APPLICABILITY: As shown for Core Protection Calculators in Table 3.3-1.

ACTION: With a Core Protection Calculator Addressable constant found to be non-conservative, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1.1 until the channel is restored to OPERABLE status.

ARKANSAS - UNIT 2 2-4 3105220N

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, C TABLE 2.2-1

=

REACTOR PROTECTIVE INSTRUMENTATMN TRIP SETPOINT LIMITS .

i

g FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

[ 1. Manual Reactor Trip Not Applicable Not Applicable

. 2. Linear Power Level - High

• a.

3)10% 4 110.711 5 Four Reactor Coolant Pumps W of RATED THERMAL POWER Operating T-M%M21.of RATED THERMAL POWER

b. Three Reactor Conlant Pumps *

• Opera ting

c. Two Reactor Coolant Pumps * *

' Operating - Same Loop m d. Two Reactor Coolant Pumps *

• i

% . Operating - Opposite Loops j 3. Logarithmic Power Level -

t liigh (1) 10 75% of RATED THERMAL POWER l 1 0.819% of RATED THERMAL POWER

4. 23G1 2370.417 Pressurizer Pressure - High  ?

1 S45. psia -

1 235h83I. psia 17t-6 I112.757

[ 5. Pressurizer Pressure - Low > 1940. psia (2) -

o _ > 16%2fi psia (2) i y 6. Containment Pressure - High 1 8.4 1 psia < 19.024' psia 3

75/

l 7. Steam Generator Pressure - Low ' 719 6/3

'[

o

> 728. psia (3) > 706:4 psia (3) 9(, .7 . W.9 fl 6 8. Steam Generator Level - Low >'_-W 5L.(4) / >' 6 613 (4)

• These values left blank pending NRC approval ofganalyses for operation with less than four reactor , coolant pumps operating. S.My I N

,_..w .a,cw _

_2 -% .va. . _ -- : - . a A rdd .w - - .

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. .J G O O O TABLE 2.2-1 (Continued [

I B REACTOR PROTECTIVE INSTRUMENTATION TRIP _SETPOINT LIMITS

.M 9

1 g J $ FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

, T, ri 5 c- 9. Local Power Density - High 1 20.3 kw/ft (5) 1 20.3 kw/ft (5)

'-iJ 5 1.2V l. W i

10. DNBR - Low ->_1 J (5) > k r (5)

~

9V.599 d 11. Steam Generator Level - High s b3.7 (4) 1'%48S% (4) s.g ,

TABLE NOTATION s

(1) Trip may be manually bypassed above 10-4% of RATED THERMAL POWER; bypass shall be automatically

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J removed when THERMAL POWER is < 10 % of RATED THERMAL POWER.

'3 m l darlyf <aned ced=chan on pressacher 1

A (2) Value may be decreased manually, to a minimum value of 100 psia, es ,,.c.,suri:cr 6 presourc reduccd, provided the margin between the pressurizer pressure and this value is maintained at pressurc is

.J 1 200 psi; the setpoint shall be increased automatically as pressurizer pressure is increased

.n until the trip setpoint is reached. Trip may be manually bypassed below 400 psia; bypass shall A

be automatically removed whenever pressurizer pressure is > 500 psia.

eturing. < plannut vehgion in sleam yneater preswee

(3) Value may be decreased manually -as eter <;enerate. pressucc is rciced, provided the margin between the steam generator pressure and this value is maintained at < 200 psi; the setpoint shall be increased automatically as steam generator pressure is increased until the trip setpoint is reached, j

(4) % of the distance between steam generator upper and lower level instrument nozzles.

, (5) As stored within the Core Protection Calculator (CPC). Calculation of the trip setpoint includes i nd processor uncertainties, and dynamic allowances. Trip may be measurement, calculational 10- g% of RATED THERMAL POWER; bypass shall be automatically removed wh manually bypassed belog%

THERMAL POWER is > 10- of RATED THERMAL POWER.

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~T~A BLE 2 . 2 - 7.

CORE P Ro Tcc.T t o r4 CALcu LATDR ADDR Ess AG LE CD t4 STAT 4 T'S

'I., TNPE I A DDRESS AB LE. Cot 4 STANT3 i

Poira r In Persrmy j A LLot0 ABLE N tA ME.E.R LABE.L D ESC R.t PT to rd VALuE 60 FCI Core coolant mass flow rate calibration constant :6 l.ts

,_ : . ' ' ;::: '. ::.:t: )

61 FC2 Core coolant mass flow rate calibration constant o.0 62 CEAf!0P CEAC/RSPT inoperable flag (~;;: -

2'- O,1, 2. or 3

. ...)_

63 TR Azimuthal tilt allowance (~;;; ' " ::: : > I . 0 2.

64 TPC Thermal power calibration constant

' ,,.. : _..._.::U - --- '- '} 2, o.9 o l

i 65 KCAL i:eutron flux power calibration constant -

>,c.85

(~ ;: '

" ~- :t'  :: :t: .)

66 DliBRPT CfGR pretrip setpoint (~;;; ' .

. .._^.. __~ . -j un r estric'recl.

67 LPDPT Local power density nretrip setpoint (~;;: :

--_,.,,.J uwrestr'icteA ,

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Agg A,usAS - UNtT 2. 2-7 P00R ORGIRL

TABLE 2. 2 - 2 ( CONTir4UE D) c.oR.e PacTec.TtoN CAL.cuL A reR. ADcdESE AR LE C.c t45TAQ TS II . TVPE Er ADDRESSABLE cot 4s. TAM rs -

PoiN T ID Percrw.i NgMBER _

LAS E.L D ESC R.t PTtod 68 BERR0 Thermal power uncertainty bias s(T ,.. !! :: M : -t %

e-: : . _ . ~. )

69 BERR1 Power uncertainty factor used,in D:'BR calculation

s. .r.- ..

70 BERR2 Power uncertainty bias used in DllBR calculation

, ,,e. 4. m.. . . . . . _ _ . . . _, -

71 BERR3 Po,ler uncertainty factor used in local power density calculation {~ r. :: . _ . Z  : ^ r + '.

72 BERR4 Power uncertainty bias used in local pcwer density calculation (T,, ~: .dl.-::: u _ _ _.

73 EOL End of life flag 'T r:

-f! . J __.:'

74 ARM 1 ijultiplier for planar radial peaking factor 75 ARM 2 llultiplier for planar radial peaking factor 7, e; .a ... .. ~ . . . ..

76 ARM 3 Multiplier "" for planar radial peaking factor

{s,_

77 ARM 4 Multiplier for planar radial peaking factor g g g. . .- --... .. .

78 ARM 5 M,ultiplier for planar radial peaking factor s . , ,. _ .. .-_ . . _. -

79 ARMS Multiplier for planar radial peaking factor ~

T, r . .: .d: : : : . 2 _ __ _' -

B3 ARl17 Multiplier- for planar radial peaking factor

u. . . . . u , -~..- .3

(

81 SCll- Shape annealing correction facter c.,,e .. . . . . . . . , - ~ . . ,

82 SCl2 Shape annealing correction factor ARR ANSAS - UNIT 2_ 2-6

TABLE 7.2-2 { CONTirJUE D)

C.O R E PROTEC.Tt or4 cat cuLATt 1 ADD f2.E'LE AR LE C.o H 5T A Q T S Ir . TVPE 3r ADDRESSABLE Cot 45.TANTS (con +Inu ed )

Poir4T ID PErGRAM N LLMBEP- LAB EL D ESC P IPTLod 83 5013 S,hape annealing correction factor

_. .. . _ s 84 SC21 Shape annealing correction factor 85 5022 Shape annealing correction factor 86 SC23 Shape annealing correction factor

s. .,. .. . - . _ . - .i 87 SC31 S,hape,, anrealing correction factor

,,-  ;; ,_- , q . _ -

88 5:32 Shape annealing correction factor 89 SC33 Shape

!= **

annealing correction factor

• .,r  %

m .,. . _-_.___ _ __ _

90 PFMLTD DiiBR penalty factor correction multiplier 1~ * - > > - - - - \

saa .- --

91 PFMLil i.P0 penalty factor correction multiplier c.,- -

.)

92 ASM2 Multiplier for CEA shadoaing factor 93 ASM3 ;1ultiplier for CEA shadoaing factor v.,- ___ _____ _ _,

ARKAbisAs - utatT 2. Z-9

TABLE '2. 2 - 2 ( CONTINUE D) c.o R E Prac REC.TtOt4 cat cuLA t ofa. ADDRESC AS Le to t4STA4 TS E. TVPE U- ADDRESSABLE cot 4s. tar 4TS bonUn ue M Poir4Y ID PErGRAM N LAMBE.R LAB EL D ESC R.t PTto M 94 ASM4

_rMultiplier

- . , - .. for CEA _shadowing , _ _ f, actor 95 ASM5 Multiplier for CEA shadowing factor

..u,_

.y 95 ASM5 Multiplier for CEA shadowing factor 1- .-

x .___ t,, --__._o 97 ASM7 Multiplier for CEA shadowing # actor

,~c- .. .... u.... .

98 CORR 1 Temperaturs. shadowing correction factor multiplier r,._- ,,_;;,.-__,u, -_-_.._.s 99 BPPCCI Coundary point power correlation coefficient r -- -- u.___..,- -__..o 100 BPPCC2 i

Coundary

_r . ..,,: ..

point power correlati.on coefficient 101 BPPCC3 Coundary point power correlation coefficient j l ,,  :: .: ___ . _ . .

)

102 BPPCC4 Boundary point power correlation coefficient t

1 kqgAgcAS,-. LidtT 7- - D 1

I _ _ _ _ , . _ - . . , - . - - - - .

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, ;s' '.G.'sI,5$1O. " t;! . 'i '

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TABLE 4.3-1 (Continued),

3, REACTOR PROTECTIVE lilSlRUMENTATI0tl OllWEILLAtlCE REQUIREMENTS

o 4 9 2

, y CilANilEL MODES IN WillCil -

5^

CilAMNEL CllAfillEL FullCTIONAL SURVEIllAllCE

' CliLCK TEST

.FUNCT10flAL UrtIT . cal. lBRATION REQtlIRED g

5 10. DilBR - Low S S(7),D(2,4), M,R(6) 1, 2 m

- c' M(ll) , It(4,5)

11. Steam Generator Level - liigh S R. M l, 2

12. Reactor Protection System j Logic N.A. fl . A . .M 1, 2 and

• l

13. Iteactor Trip Breake'rs fl . A . N.A. M 1,2 and

• l 14. Core Protection Calculators 5, W$) D(2,4),R(4,5) M,R(6) 1, 2 TS 15 CEA Calculators S R M,R(6) 1, 2 O Y' C:3

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(" , TABLE 4.3-1 (Continued)

TABLE NOTATION With reactor trip breakers in the closed positicn and the CEA drive system capaole_of CEA withdrawal.

(1) -

If not performed in previous 7 days.

(2) -

Heat balance only (CHANNEL FUNCTIONAL TEST not included),

above 15". of RATED THERMAL POWER; adjust the Linear Power

-- Level signals and the CPC addressable constant multipliers .

to make the CPC aT power and CPC nuclear' power calculations agree with the calorimetric calculation if absolute difference is > 2".. D'Jring PHYSICS TESTS, these daily calibrations may be suspended provided these calibrations are performed upon

, reaching each major test pcwer plateau and prier to proceeding

to the next major test pcwer plateau.

(3) -

Above 15% of RATED THERMAL POWER, verify that the linear power subchannel gains of the excore detectors are consistent with

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the values used to establish the shape annealing natrix elements in the Core Protection Calculators.

(J) -

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(' .,.

(5) -

After each fuel loading and prior to exceeding 70% of RATED THERMAL PCWER, the incere detectors.shall be used to determine the shace annealing matrix elements and the Ccre Protection Calculators'shall use tnese elements.

(5) -

This CHANNEL FUNCTIONAL TEST shall include the injection of

- simulated precess signals into tne chanr.el as close to the

" sensors as practicable to verify OPERABILITY including alarm N and/or trip functions.

k (7) - Above 70". of RATED THERMAL POWER, verify that the total RCS flow rate as indicated.by each CPC is less than or equal to

.; the actdal RCS tota'l ficw rate determined by either using the

1. ' 'eactor coolant pumo differential pressure instrumentation 1 (conservatively ccmcensated for measuremer.: uncertainties) or by calorimetric calculations (censervatively compensated for

~

measurement uncertainties) and if necessary, adjust the CPC 1 addressable constant ficw coefficients such that each CPC indicated flow is less than or equal to tne actual flow rate.

, 'The ficw measurement uncertainty may be included in the BERR1 term in the CPC and is equal to or greater than 4"..

(3) - Abcve 70" of RATED THERMAL POWER, verify that the total RCS e flow rate as indicated by eacn CPC is less than or equal to the actual RCS total ficw rate determined by calorimetric calculations (conservatively ccmcensated for measurerent uncertainties).

((O h arre et Va.h c-C adbeccetM ud-huds ( See Ta.lo(c 2 2-2) sham W. VeriUe d -to be, 'msdnQtd iw cc.ch CPERABLE CPC..

ARKANSAS - UNIT 2 3/4 3-9

__ _ - - ~ -

SAFETY-LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES

a. 'RCS Cold Leg Temperature-Low > 465*F

b. RCS Cold Leg Temperature-High. ;c605*F

c. Axial Shape Index-Positive Not more positive than +0.6

d. Axial Shape Index-Negative Not more negative thanl-0.6

e. Pressurizer Pressure-Low jt 1705 psia

f. Pressurizer Pressure-High jt2400 psia

g. Integrated radial Peaking Factor-Low > 1.28

h. Integrated Radial Peaking Factor-High 504.2"

i. Quality Margin-Low jt0 Steam Generator Level-High The Steam Generator Level-High trip is provided to protect the turbine from excessive moisture carry over. Since the turbine is automatically tripped when the reactor is tripped, this trip provides a reliable means for providing protection to the turbine from excessive moisture' carry over. This trip's setpoint does not-correspond to a. Safety Limit and no credit was taken in the accident analyses for operation of this trip.

Its functional capability at the specified trip setting is required to enhance the overall reliability of the Reactor Protection System.

2.2.2 CPC Addressable Constants The Core Protection Calculator (CPC) addressable constants are provided to allow calibration of the CPC system to more accurate indications such as.colorimetric measurements for power level and RCS flowrate and incore detector signals for axial flux shape, radial peaking factors and CEA deviation penalties. Other CPC addressable constants allow penalization of the calculated DNBR and LPD values based on measurement uncertainties or inoperable equipment. Administrative controls on changes and periodic checking of addressable constant values (see also Technical Specirica-tions 3.3.1.1 and 6.8.1) ensures that inadvertent misloading is un-likely.

1 B 2-7

ADMINISTRATIVE CONTROLS

~ 6.7- SAFETY LIMIT VIOLATION ,

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6.7.1 The following. actions shall be taken in'the event a Safety Limit is violated:

a. The unit shall be placed in at least HOT STANDBY within one hour.

. b. The Safety Limit violation shall be reported to the Commission, the Director, Nuclear operations and to the SFC within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the PSC. This report shall describe (1) applicable circumstances preceding the

-violation, (2) effects of the violation upon facility components, systems or structures and (3) corrective action taken to prevent recurrence.

d. The Safety Limit Violation Report shall be submitted to the Commission, the SRC and the Director, Nuclear Opera-tions within 14 days of the violation.

6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented and

, maintained covering the activities referenced below:

a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February 1978.

b. Refueling operations.

c. Surveillance and test activities of safety related equipment.

d. Security Plan implementation.

r e. Emergency Plan implementation. ,

f. Fire Protection Program implementation.

g. Modification of Core Protection Calculator (CPC) Ad-dressable Constants NOTE: Modification to the CPC addressable constants based on information obtained through the Plant l

Computer - CPC data link shall not be made without prior approval of the Plant Safety Committee.

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f 6- 13 L _ __ _ _ _ _ _

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-ADMINISTRATIVE CONTROLS (Cont.)

Each procedure of-6.'8.1 above, and' changes thereto, shall be

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6.8.2 reviewed by the PSC and approved by the General Manager prior to. implementation and reviewed periodically _as set forth in administrative procedures.

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

POWER DISTRIBUTION LIMITS

( AXIAL SHAPE INDEX LIMITING CONDITION FOR OPERATION 3.2.7 The core average AXIAL SHAPE INDEX (ASI) shall be maintained within the following limits:

a) COLSS OPERABLE

-0.28 < ASI < + 0.28 b) COLSS OUT OF SERVICE (CPC)

-0.20 < ASI < +0.20 APPLICABILITY: MODE 1 above 20% of RATED THERMAL POWER ACTION:

With the core average AXIAL SHAPE INDEX (ASI) exceeding its limit, restore the ASI to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 20% of RATED THEPFAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, b

SURVEILLANCE REQUIREMENTS 4.2.6 The core average AXIAL SHAPE INDEX shall be detennined to be within its limits i

at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> using the COLSS or any OPERABLE Core Protection Calculator channel.

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W $ee. Spec's ed Te st E v c e.p6on 3. t O . 2.

. (

ARKANSAS - UNIT 2 3/4 2-12

SPECIAL TEST EXCEPTIONS GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The group height, insertion and power distribution limits of Specifications 3.1.1.4, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.2, 3.2.3, 3.2.7 and the Minimum Channels OPERABLE requirement of Functional Unit 15 of Table 3.3-1 may be suspended during the performance of PHYSICS TESTS provided.

a. The THERMAL POWER is restricted to the test power plateau which shall not exceed 85% of RATED THERMAL POWER, and

'b. The limits of Specification 3.2.1 are maintained and determined as specified in Specification 4.10.2.2 below.

APPLICABILITY: During startup and PHYSICS TESTS.

ACTION:

With any of the limits of Specification 3.2.1 being exceeded while any of the above requirements supended, either:

a. Reduce THERMAL POWER sufficiently to satisfy the requirements of Specification 3.2.1, or

b. Be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined at least once per hour during PHYSICS TESTS in which any of the above requirements are suspended and shall be verified to be within the test power plateau.

4.10.2.2 The linear heat rate shall be determined to be within the limits of Specification 3.2.1 by monitoring it continuously with the Incore. Detector Monitoring System pursuant to the requirements of Specifications 4.2.1.3 and 3.3.3.2 during PHYSICS TESTS above 5% of RATED THERMAL POWER in which any of the above requirements are suspended.

3/4 io-2

INSTRUMENTATION SURVEILLANCE REQUIREMENTS (Continued)

2. With 120 volts AC (60 Hz) applied for at-least 30 seconds across the input, the reading on the output does not.

exceed 8 volts DC.

.b. For the optical isolators: Verify that the_ input to output insulation resistence is greater than 10 megohms when tested using a megohmeter on the 500 volt DC range.

4.3.1.1.5 The Core Protection Calculator System shall be determined OPERABLE at lease once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that less than three auto restarts have occurred on each calculator-during the past 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.3.1.1.6 The Core Protection Calculator System shall be subjected to a CHANNEL FUNCTIONAL TEST to verify OPERABILITY within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of receipt of a valid High CPC Room Temperature alarm.

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UNITED STATES g [  %'h NUCLEAR REGULATORY COMMISSION 2'IDO

. p WASHINGTON. D. I' 20555

'+,***** p April 10, 1981 C.\ h%h .

Docket No. 50-368

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nm - Aquss+ Gr oddlemo) in6. oS ,

hh)b i CAiCl6 b bEId 3@d-Mr. William Cavanaugh, III ,

Senior Vice President 2 ,, w Energy Supply Department APR 161981 Arkansas Power & Light Company P. O. Box 551 Little Rock, Arkansas 72203 ARVANSAS POWER & UGHT CO.

GENE. OAT 10N & CONSTRUCTION

Dear Mr. Cavanaugh:

The staff has reviewed your February 20,19817ild March 5,1981 subnittals on the ANO-2 Cycle 2 reload Er.d has identified a need for additional information as set forth in the enclosure.

Please contact us if you have questions regarding the items noted in the enclosure.

Sincerely, 7

Robert A. Clark, Chief Operating Reactors Branch #3 Division of Licensing

Enclosure:

As stated

-ull y L APR i c w

?s$s$*uflyE T 3 /6 11 6 M B M 1 ( ~/y y

• 48 Describe the bases for the change from less than or equal to (3/4 7-8) 0.10 micro curies oer gram dose equivalent I-131 to less than or equal to 0.046' micro curies per gram.

61 The recent problems you refer to with the elect'r ic driven feed-(B3/4 7-2) water pump suggest that the pumping capability, although cap-able of meeting revised safety analysis considerations, may have been reducted somewhat. Outline your plans for evaluating this matter and provide a schedule for reporting to the staff the results of your evaluation and correctivo actions to be taken as required.

PART II - INSTRUMENTATION AND CONTROLS SYSTEM Q-l . The staff feels that AP&L Co., should propose te:hnical specific- S**-Pri'5eg ations to assure that the CPC is not considered operaole when environmental conditions including cyclic or ramped temperature T' E - 4'3 l

• l ' (8 fluctua t. ion exceed those for which the CPC has been cualifiH.

Provide justification for the environmental limits proposed.

Q-2. Table 3-4 of CEN 147(s)-P contains upper and lower proposed allowed bounds on addressable constants. These bounds as

- curre.ntly proposed would restrict the values of addressable constants entered into the CPC to avoid only very gross errors.

Other, smaller, yet unacceptable values could be entered. For example, a negative value of a diagonal element of the shape annealing correction matrix'does not seem justified and such -

values should be rejected by the computer. Furthemore, there may be values of addressable constants within the current pro-posed bounds which if entered could lead to violation of DNBR or LPD limits even when the CPC is otherwise functioning pro- -

perly.

Therefore, please adopt more restrictive bounds on the address-able constants to assure that values may not be entered which are physically unrealistic or which could lead to violation of DNBR or LPD limits even when the CPC is otherwise functioning l properly.

l

' Provide a commitment to so modify CPC addressable constant limits at the next CPC software ~ change but not later than six ..

months from the date of this letter.

Q-3. The staff feels that AP&L Co., should propose technical specifications to assure that (a) plant procedures shall See proposed be in effect to control modifications to CPC addressable 'T . S . G . O . l . 9 constants (b) these procedures are consistent with Approved Physics and Thermal Hydraulic Methods; the approved methods should be referenced in the bases to the Technical Specific-ations (c) CPC Addressable constants and thei- allowed ranges see p ro posed (i.e., upper and lower bounds) are identified in the Teci.nical T. s. 2. 2. 2.

Specifications (d) values of Addressable Constants outside the allowed range are not to be entered without approval of the Plant Safety Committee (e) An independent verification l

• ?

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see Pr shall be conducted to confirm that Addressable Constant Modifications O*""S*f{'5

• have been made as approved by the Plant Safety Committee or the rabic 4,3.-

Encineering Staff (whichever is applicable) (f) Modifications to the CPC Addressable Constants based on information obtained through ), see. Propos e the Plant Computer Data Links shall not be made without approval L T. S . G. B. I. cf of the Plant Safety Committee. -s PART III - OTHER ISSUES -

Q-1. Your letter of August 29, 1980 requested an extension frem 12 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for setting the pressurizer code safety valves _ during Mode 3. The following information is needed to enable our review.

a) Was the subject testing performed as -part of the ANO-2 ASME Code Section XI inservice testing and _ inspection program?

b) How many tests have been conducted on these valves to date and what was the time required to do- each of these tes ts ?

c) How does AP&L Co's., experience with the tesuing of these valves compare to general industry practice?

d) During the testing are both valves rendered inoperable at the same time?

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