ML20029C979

Amend 94 to License NPF-38 Re Changes to App a TS by Moving RTS & ESFAS Response Time Limits from TS to Updated FSAR

ML20029C979
Person / Time
Site:	Waterford
Issue date:	04/22/1994
From:	Beckner W Office of Nuclear Reactor Regulation
To:
Shared Package
ML20029C980	List: ML20029C979 ML20029C981
References
NUDOCS 9405030151
Download: ML20029C979 (9)

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UNITED STATES

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NUCLEAR REGULATORY COMMISSION g

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W ASHINGTON, D.C. 205SS-4001 ENTERGY OPERATIONS. INC.

DOCKET NO. 50-382 WATERFORD STEAM ELECTRIC STATION. UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 94 License No. NPF-38 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Entergy Operations, Inc. (the licensee) dated February 14, 1994, complies with the standards and reouirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

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9405030151 940422 DR ADOCK 0500 2

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Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Facility Ope. rating License No. NPF-38 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 94, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.

The licensee shall operate the facility-in accordance with the Technical Specifications and the Envircnmental Protection Plan.

3.

This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION f

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William D. Beckner, Director Project Directorate IV-1 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

April 22, 1994 l

4 ATTACHMENT T0 LICENSE AMENDMENT NO. 94 TO FACILITY OPERATING LICENSE NO. NPF-38 DOCKET NO. 50-382 Replace the following pages of the Appendix A Technical Specifications with the attached pages.

The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change. The corresponding overleaf pages are also provided to maintain document completeness.

REMOVE PAGES INSERT PAGES XX XX 3/4 3-1 3/4 3-1 3/4 3-8 3/4 3-8 3/4 3-9 3/4 3-13 3/4 3-13 3/4 3-22 3/4 3-22 3/4 3-23 3/4 3-24 s

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INDEX LIST OF FIGURES FIGURE PAGE 3.1-0 SHUTDOWN MARGIN AS A FUNCTION OF COLD LEG TEMPERATURE............................

3/4 1-3a 3.1-1 REQUIRED STORED BORIC ACID VOLUME AS A FUNCTION OF CONCENTRATION......................................

3/4 1-13 3.1-1A REQUIRED POWER REDUCTION AFTER SINGLE CEA DEVIATION 3/4 1-20a 3.1-2 CEA INSERTION LIMITS VS THERMAL POWER..............

3/4 1-27 3.1-3 PART LENGTH CEA INSERTION LIMIT VS THERMAL POWER...

3/4 1-28a 3.2-1 ALLOWABLE PEAK LINEAR HEAT RATE VS Tc..............

3/4 2-2 3.2-1A ALLOWABLE PEAK LINEAR HEAT RATE VS Tc FOR COLSS OUT OF SERVICE.........................................

3/4 2-2A 3.2-2 DNBRMARGINOPERATINGLIMITBASEDONCOREPROTECTkON CALCULATORS (COLSS OUT OF SERVICE, CEACs OPERABLE)..

3/4 2-8 3.2-3 DNBR MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATORS (COLSS OUT OF SERVICE CEACS IN0PERABLE).........................................

3/4 2-9 3.4-1 DOSE EQUIVALENT I-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE PRIMARY COOLANT SPECIFIC ACTIVITY >1.0 pCi/ GRAM DOSE EQUIVALENT I-131.......

3/4 4-27 3.4-2 REACTOR COOLANT SYSTEM PRESSURE / TEMPERATURE LIMITATIONS FOR 0-8 EFFECTIVE FULL POWER YEARS (HEATUP)...........................................

3/4 4-30 3.4-3 REACTOR COOLANT SYSTEM PRESSURE / TEMPERATURE LIMITATIONS FOR 0-8 EFFECTIVE FULL POWER YEARS (C00LDOWN).........................................

3/4 4-31 4.7-1 SAMPLING PLAN FOR SNUBBER FUNCTIONAL TEST..........

3/4 7-26 5.1-1 EXCLUSION AREA.....................................

5-2 5.1-2 LOW POPULATION 20NE................................

5-3 5.1-3 SITE BOUNDARY FOR RADI0 ACTIVE GASEOUS AND LIQUID EFFLUENTS...................................

5-4 6.2-1 0FFSITE ORGANIZATION FOR MANAGEMENT AND TECHNICAL SUPP0RT.................................

6-3 6.2-2 PLANT OPERATIONS ORGANIZATION......................

6-4 WATERFORD - UNIT 3 XIX AMENDMENT NO. 13, 27

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INDEX LIST OF TABLES I

TABLE PA.GE 1.1 FREQUENCY N0TATION......................................

1-9 1.2 OPERATIONAL M0 DES........................................

1-10 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS..................................................

2-3 2.2-2 CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS........

2-5 MONITORING FREQUENCIES FOR BORON DILUTION DETECTION 3.1-1 K,,, > 0.98.............................................

3/4 1-17 3.1-2 0.98 2 K,,,

> 0.97......................................

3/4 1-17a 3.1-3 0.97 2 K,,, > 0.96......................................

3/4 1-17b 3.1-4 0.96 2 K,,,

> 0.95......................................

3/4 1-17c 3.1-5 K,,, s 0.95.............................................

3/4 1-17d 3.3-1 REACTOR PROTECTIVE INSTRUMENTATION......................

3/4 3-3 4.3-1 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................

3/4 3-10 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.........................................

3/4 3-14 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES.............................

3/4 3-19 I

4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS...............

3/4 3-25 3.3-6 RADIATION MONITORING INSTRUMENTATION....................

3/4 3-29 4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...............................

3/4 3-32 3.3-7 SEISMIC MONITORING INSTRUMENTATION......................

3/4 3-36 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...............................

3/4 3-37 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION...............

3/4 3-39 WATERFORD - UNIT 3 XX AMENDMENT N0. 9 94 7

3 /4. 3 INSTRUMENTATIDH 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protective instrumentation channels and bypasses of Table 3.3-1 shall be OPERABLE.

APPLICABILITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

SURVEILLANCE RE0VIREMENTS 4.3.1.1 Each reactor protective instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-1.

4.3.1.2 The logic for the bypasses shall be demonstrated OPERABLE prior to each reactor startup unless performed during the preceding 92 days. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation.

4.3.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be demonstrated to be within its limit at least once per 18 months.

Neutron detectors are exempt from response time. testing.

Each test shall l

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include at least one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the " Total No. of Channels" column of Table 3.3-1.

4.3.1.4 The isolation characteristics of each CEA isolation amplifier and each optical isolator for CEA Calculator to Core Protection Calc'ulator data.

transfer shall be verified at least once per 18 months during the shutdown per the following tests:

a.

For the CEA position isolation amplifiers:

1.

With 120 volts AC (60 Hz) applied for at least 30 seconds across the output, the reading on the input does not exceed 0.015 volts DC.

WATERFORD - UNIT 3 3/4 3-1 AMENDMENT NO. 94 i

L 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-15.0 volts DC.

p b.

For the optical isolators:

Verify that the input to output insulation resistance is greater than 10 megohms when tested.using a megohmmeter on the 500 volt DC range.

4.3.1.5 The Core Protection Calculator System and the Control Element Assembly Calculator System shall be determined OPERABLE at least 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 />.

4.3.1.6 The Core Protection Calculator System shall be subjected to' a CHANNEL

-i FUNCTIONAL TEST to verify OPERABILITY within 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />s-of. receipt of a High CPC Cabinet Temperature alarm.

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t WATERFORD - UNIT 3 3/4 3-2

'4-TABLE 3.3-1 (Continued) 41 ACTION STATEMENTS-

- 2.

Within 4-hours:

a)

All full-length and part-length CEA groups are withdrawn to and subsequently maintained at the

" Full Out" position,.except during surveillance testing pursuant to'the. requirements of Specification 4.1.3.1.2 or for control when.

CEA group 6 may be inserted no further than 127.5 inches withdrawn.

b)

The "RSPT/CEAC Inoperable" addressable constant in the CPCs is set to the inoperable status.

c)

The Control Element Drive-Mechanism Control.

System (CEDMCS) is placed.in and subsequently i

maintained in the "Off" mode except during CEA

. group 6 motion permitted by a) above, when the CEDMCS may be operated in either the Manual Group" or " Manual' Individual" mode.

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At least'once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, all full-length and part-q length CEAs are verified. fully withdrawn except

]j during surveillance testing pursuant to Specifica-tion;4.1.3.1.2 or during insertion of.CEA. group 6 as

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permitted.by 2.a) above, then verify at least once

' j per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.that the inserted CEAs are ' aligned within 7 inches (indicated position) of all other CEAs in its group.

ACTION 7 With three or more auto restarts of one non-bypassed calculator during a 12-hour interval, demonstrate calculator-OPERABILITY-by performing a CHANNEL FUNCTIONAL TEST within the next

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24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

l ACTION 8 With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or open the reactor trip breakers within the next hour.

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WATERFORD UNIT 3 3/4 3-7 1

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TABLE 3.3.2 has been deleted.

4 WATERFORD - UNIT 3 3/4 3-8 AMENDMENT NO. 12,40,04 Next page is 3/4 3-10

t0 TAB'LE 4.3-1 REACTOR PROTTCTIVE INSTPUMENTAT10N SURVEILLANCE REQUIREMENT

_1 e

c CHANNEL MODES FOR WHICH FtWCTIONAL UNIT CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE CHECK CALIBRATION TEST

_IS REQUIRED w

1.

Manual Reactor Trip H.A.

N. A.

R and S/U(1) 1, 2, 3*, 4*, 5*

7 Linear Power Level - High 5

0(2,4),M(3,4), Q 1, 2 Q(4) 3.

Logarithmic Power Level - High S

R(4)

Q and S/U(1) 27, 3, 4, 5 4

Pressurizer Pressure - High 5

R Q

1, 2 5.

Pressurizer Pressure - Low 5

R Q

1, 2 s

[

6. -Containment Pressure - High 5

R Q

1, 2 M

7 Steam Generator Pressure - Low 5

R Q

1, 2 i

8.

Steam Generator Level - Low S

R Q

1, 2 9.

Local Power Density - High S

D(2,4), R(4,5) Q, R(6) 1, 2 10.

DNBR - Low S

l S(7), 0(2,4),

Q,R(6) 1, 2 l

M(8), R(4,5) 11.

Steam Generator Level - High S

R Q

1, 2 9 12.

Reactor Protection System y

logic N.A.

H.A.

Q and S/U(1) 1, 2, 3 *, 4 *, 5

• 5 5

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INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION' SYSTEM INSTRUMENTS 11DH LIMITING CONDITION FOR OPERATION f

3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and bypasses shown in Table 3.3-3 shall be OPERABLE with their. trip setpoints set consistent with the values shown in the Trip Setpoint columnlof Table 3.3-4.

l-APPLICABILITY: As shown in Table 3.3-3.

ACTION:

With an ESFAS instrumentation channel trip setpoint less conservative a.

than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION.

3 requirement of Table 3.3-3 until the channel. is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value, b.

With an ESFAS instrumentation channel inoperable, take the ACTION shown in Table 3.3-3.

SURVEILLANCE RE0VIREMENTS 4.3.2.1 Erich ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-2.

4.3.2.2 The logic for the bypasses shall be demonstrated OPERABLE during the at power CHANNEL ' FUNCTIONAL TEST of channels affected by bypass operation..The:

total bypass function shall be demonstrated OPERABLE at-least once ~per 18 months during CHANNEL CALIBRATION testing of each channel affected.by bypass-operation.

4.3.2.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months.

Each test shall include at least one channel per function such that all channels 'are tested at least once every N times 18 months where N is the total number of-redundant channels in 'a specific ESFAS function as shown in the " Total No. of Channels" Column of Table 3.3-3.

3 WATERFORD - UNIT 3 3/4 3-13 AMENDMENT NO. 94 w

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TABLE 3.3-3 C

[-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION m

3 E

MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION C5 1.

SAFETY INJECTION (SIAS)

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a.

Manual,(Trip Buttons) 2 sets of 2 1 set of 2 2 sets of 2 1,2,3,4 12 b.

Containment Pressure -

High 4

2 3

1,2,3 13*, 14*

c.

Pressurizer Pressure -

Low 4

2 3

1, 2, 3(a) 13*, 14*

d.

Automatic Actuation -

Logic 4

2 3

1,2,3 12 ca 2.

CONTAINMENT SPRAY (CSAS) 1 a.

Manual (Trip Buttons) 2 sets of 2 1 set of 2 2 sets of 2 1,2,3,4 12 T

b.

Containment Pressere --

High - High 4

2(b) 3 1,2,3 13*, 14*

c.

Automatic Actuation Logic 4

2 3

1,2,3 12 3.

CONTAINMENT ISOLATION (CIAS) a.

Manual CIAS (Trip Buttons) 2 sets of 2 1 set of 2 2 sets of 2 1,2,3,4 12 b.

Containment Pressure -

High 4

2-3 1,2,3 13*, 14*

c.

Pressurizer Pressure -

Low 4

2 3

1, 2, 3(a) 13*, 14*

d.

Automatic Actuation Logic 4

2 3

1,2,3 12 4

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TABLE 3.3-4 (Continued)

TABLE NOTATIONS (1) Value may be decreased manually, to a minimum of 100 psia, as pressurizer pressure is reduced, provided the margin between the pressurizer and this value is maintained at less than or equal to 400 psi; the setpoint shall be increased automatically as pressurizer pressure is increased until the trip setpoint is reached.

Trip may be manually bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer is greater than or equal to 500 psia.

(2) Value may be decreased manually as steam generator pressure is reduced, provided the margin between the steam generator pressure and this value is maintained at less than or equal to 200 psi; the setpoint shall be increased automatically as steam generator pressure is increased until the trip setpoint is reached.

(3) % of this.dictence between steam generator upper and lower level instrument

-mn i e s.

(4) Requires corresponding permissive trip signal of item 7.c., 7.d., or 7.e.

to actuate EFAS.

(5) Recuires corresponding EFAS trip to actuate control valves.

I WATERFORD - UNIT 3 3/4 3-21 AMENDMENT NO.19

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WATERFORD - UNIT 3 3/4 3-22 AMENDMENT NO. 74,75, 94 Next page is 3/4 3-25

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