Text

Proposed Tech Spec Re Cooling & Protection of Fuel in Reactor Vessel After Removal of All Fuel
	ML20081B157
Person / Time
Site:	Hatch
Issue date:	10/20/1983
From:	; GEORGIA POWER CO.
To:	;
Shared Package
ML20081B151	List: ML20081B148; ML20081B157;
References
	TAC-52483, NUDOCS 8310270334
	Download: ML20081B157 (26)
	v • d • e

'.

ENCIOSURE 1 NRC DOCKET 50-366 OPERATING LICENSE NPF-5 EDNIN I. HA'IGI NUCLEAR PIANI' UNIT 2 REQUEST 'IO AMEND TECHNICAL SPECIFICATIONS

'Ihe proposed changes to the Technical Specifications (Appendix A to Operating License NPF-5) would be incorporated as follows:

Renove Page Insert Page 2-2 2-2 3/4 1-8 3/4 1-8 3/4 1-9 3/4 1-9 3/4 1-11 3/4 1-11 3/4 1-18 3/4 1-18 3/4 3-2 3/4 3-2 3/4 3-3 3/4 3-3 3/4 3-5 3/4 3-5 3/4 3-12 3/4 3-12 3/4 3-15 3/4 3-15 3/4 3-21 3/4 3-21 3/4 3-26 3/4 3-26 3/4 3-31 3/4 3-31 3/4 3-38 3/4 3-38 3/4 3-39 3/4 3-39 3/4 3-41 3/4 3-41 3/4 3-42 3/4 3-42 3/4 3-48 3/4 3-48 3/4 3-60 3/4 3-60 3/4 5-7 3/4 5-7 3/4 9-3 3/4 9-3 3/4 9-5 3/4 9-5 3/4 9-20 3/4 9-20 l

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l i

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8310270334 831020 PDR ADOCK 05000366 PDR l

P L

SAFErY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFEIY LIMITS (Continued)

REMMDR VESSEL WATER LEVEL 2.1.4

%e reactor vessel water level shall be above the top of the active irradiated fuel.

APPLICABILITY: CONDITIOtE 3, 4 and 5*

ACTION:

With the reactor vessel water level at or below the top of the active irradiated fuel, manually initiate the low pressure BCCS to restore the reactor vessel water level, after depressurizing the reactor vessel, if required.

l

Not applicable while all fuel is removed from the reactor vessel.

l l

l HA'ICH - UNIT 2 2-2 1

l r

REACTIVI'IY CONTRG, SYSTD4S CCNIROL BOD SCRAM ACCGUIA'KRS LIMITING CX)NDITION FOR OPERATION 3.1.3.5 All control rod scram accumulators shall be OPERABLE.

APPLICABILITY: CONDITIONS 1, 2 and 5*(a),

l ACTION:

a.

In CONDITION 1 or 2 with one control rod scram accumulator inoperable, the provisions of Specification 3.0.4 are not applicable and operation may continue, provided that within 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s:

1.

We inoperable accumulator is restored to OPERABLE status, or 2.

The control rod associated with the inoperable accumulator is declared inoperable and the requirements of Specification 3.1.3.1 are satisfied.

Otherwise, be in at least HOT SHUIDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

In CONDITIOM 5*

with a withdrawn control rod scram accumulator inoperable, fully insert the affected control rod and electrically disarm the directional control valves or close the withdraw isolation valve within one hour. %e provisions of Specification 3.0.3 are not applicable.

SURVEILINCE REDUIRFNNPS 4.1.3.5

%e control rod scram accumulators shall be determined OPERABLE:

[

a.

At least once per 7 days by verifying that the pressure and leak detectors are not in the alarmed condition, and b.

At least once per 18 months by performance of a:

1.

CHANNEL EWCTIONAL TEST of the leak detectors, and 2.

CHANNEL CALIBRATION of the pressure detectors to alarm at 955 +~

15 psig.

At least the accumulator associated with each withdrawn control rod. Not l

applicable to control rods removed per Specification 3.9.11.1 or 3.9.11.2.

(a) Not applicable while all fuel is removed from the reactor vessel.

HA'IGI - UNIT 2 3/4 1-8 1

I l

~

REACTIVITY C0tfIROL SYSTEMS CCNIROL RCD DRIVE (IXJPLING LIMITING (DDITION FOR OPERATION 3.1.3.6 All control rods shall be coupled to their drive mechanisms.

APPLICABILITY: CONDITIOtB 1, 2 and 5*(a),

l ACTION:

a.

In CONDITION 1 or 2 with one control rod not coupled to its associated drive mechanism, the provisions of Specification 3.0.4 are not applicable and operation may continue provided; 1.

If permitted by the Rm and RSCS, the control rod drive mechanism is inserted to accomplish recoupling and recoupliry is verified by demonstrating that the control rod will not go to the overtravel position, or 2.

If recoupling is not accomplished on the first attenpt or if not permitted by the Rm or RSCS, the control rod is declared inoperable and fully

inserted, and the requirements of Specification 3.1.3.1 are satisfied.

b.

In CONDITION 5*,

with a withdrawn control rod not coupled to its associated drive mechanism, within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s:

1.

Insert the control rod to acconplish recoupling and verify recoupling by demonstrating that the control rod will not go to the overtravel position, or 2.

If recoupling is not accomplished, fully insert the control rod and either electrically disarm the control rod or close the withdraw isolation valve.

3.

'Ihe provisions of Specification 3.0.3 are not applicable.

At least each withdrawn control rod. Not applicable to control rods removed per Specification 3.9.11.1 or 3.9.11.2.

(a)Not applicable while all fuel is removed from the reactor vessel.

,l HNIG - UNIT 2 3/4 1-9

REACTIVITY CONTROL SYSTFMS CONIRCL ROD ICSITION INDICATION LIMITIE COEITION FOR OPERATION 3.1.3.7 All control rod reed switch position indicators shall be OPERABLE.

APPLICABILITY: 00NDITIOts 1, 2 and 5* (a),

l

/CTION:

a.

In 00EITION 1 or 2:

1.

With one or more control rod reed switch position indicators inoperable, except for the " Full-in" or " Full-out" indicators, the provisions of Specification 3.0.4 are not applicable and operation may continue, provided that within one hour:

a)

The position of the control rod is determined by an alternate method, or b)

'Ite control rod is moved to a position with an OPERABLE reed switch position indicator, or c)

The control rod is declared inoperable and the requirements of Specification 3.1.3.1 are satisfied; otherwise, be in at least HOP SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2.

With one or more control rod reed switch " Full-in" and/or

" Full-out" position indicators inoperable, the affected control rod may be bypassed in the Rod Sequence Control System, the provisions of Specification 3.0.4 are not applicable and operation may continue, provided; a)

The actual control rod positon is known and, b)

The affected control rod is moved to the correct position in the proper sequence, b.

In CONDITION 5* with a withdrawn control rod reed switch position indicator inoperable, move the control rod to a position with an OPERABLE reed switch position indicator or fully insert the control rod. The provisions of Specification 3.0.3 are not applicable.

At least each withdrawn control rod.

Not applicable to control rods removed per Specification 3.9.11.1 or 3.9.11.2.

(a)Not applicable while all fuel is removed from the reactor vessel.

l HA'IGI - UNIT 2 3/4 1-11

REACTIVITY CONTRG, SYS'IDE 3/4.1.5 STA!OBY LIQUID CONTROL SYSITM LIMITING COMITION FOR OPERATION 3.1.5 The standby liquid control system shall be OPERABLE with:

a.

An OPERABLE flow path from the storage tank to the reactor core containing two pumps and two inline explosive injection valves, and b.

'Ihe contained solution concentration and the solution temperature are within the Operating Range of Figure 3.1.5-1.

APPLICABILITY: 00EITIONS 1, 2, and 5*(a),

l ACTION:

a.

In CONDITION 1 or 2:

1.

With one punp and/or one explosive valve inoperable, restore the inoperable punp and/or explosive valve to OPERABLE status within 7 days or be in at least HOT SIUIDODN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2.

With the standby liquid control system inoperable, restore the system to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOP SIUIDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

In CONDITION 5*:

1.

With one punp and/or one explosive valve inoperable, restore the inoperable punp and/or explosive valve to OPERABLE status within 30 days or fully insert all insertable control rods within the next hour.

1 I

2.

With the standby liquid control system inoperable, fully insert all insertable control rods within one hour.

3.

The provisions of Specification 3.0.3 and 3.0.4 are not applicable,

With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.11.1 or 3.9.11.2.

(a)Not applicable while all fuel is removed from the reactor vessel.

]

HA'IGI - UNIT 2 3/4 1-18

TABLE 3.3.1-1 REAC'IOR PROTBCPION SYSTEM INS'IREDENTATION I

APPLICABLE MINIl4N NLNBER OPERATIONAL OPERABLE CHANNELS E

Ft1CTIONAL UNIT CONDITIOtB TER 'IRIP SYSTEM (a)

ACTION U

w 1.

Intermediate Range Monitors:

(2C51-K601 A, B, C, D, E, F, G, H) a.

Neutron Flux - High 2(c),5 (b) (1) 3 1

l 3, 4 2

2 b.

Inoperative 2, 5(b) (1) 3 1

l 3, 4 2

2 2.

Average Power Range Monitor (2C51-K605 A, B, C, D, E, F) l a.

Neutron Flux - Upscale, 15%

2, 5(1) 2 1

g}

b.

Flow Referenced Simulated Thermal Power - Ipscale 1

2 3

w c.

Fixed Neutron Flux -

N tpscale, 118%

1 2

3 l

d.

Inoperative 1, 2, 5(1) 2 4

e.

Downscale 1

2 3

f.

LPRM 1, 2, 5 (1)

(d)

NA l

3.

Reactor Vessel Steam Dome Pressure -

High (2B21-N023 A, B, C, D) 1, 2(e) 2(j, 2B21-N045-A, B, C, D) 5 4.

Reactor Vessel Water Ievel -

Iow (2B21-N017 A, B, C, D) 1, 2 2(j, 2B21-N024-A, B aM 5

2B21-N025-A, B) 5.

Main Steam Line Isolation Valve -

6 Closure (NA) 1(f) 4 3

[

6.

Main Steam Line Radiation - High 1, 2 (e) 2 6

(2Dll-K503 A, B, C, D) 7.

Drywell Pressure - High 1, 2(9) 2 5

(2C71-N002 A, B, C, D)

TABLE 3.3.1-1 (Continued) g REACIOR PROIECTION SYSTEM INS'IRENENI'ATION d

i APPLICABLE MINIMM NLNBER OPERATIO@l OPERABLE CHAbNELS h

FUNCTIONAL UNIT CONDITIONS PER 'IRIP SYSTEM (a)

ACI' ION 8

w 8.

Scram Discharge Volume Water Level - High (2Cll-N013 A, B, C, D) 1, 2, 5 (h) (1) 2 4

l 9.

Turbine Stop Valve - Closure (NA) 1(i) 4 (k) 7

10. Turbine Control valve Fast Closure, 1(i) 2(k) 7 Trip Oil Pressure - Iow j

(2C71-N005 A, B, C, D) i

11. Reactor Mode Switch in Shutdown Position (NA) 1, 2, 3, 4, 5(1) 1 8

l

12. Manual Scram (NA) 1, 2, 3, 4, 5(1) 1 9

l P

i l

1

TABLE 3.3.1-1 (Continued)

RE7CIOR PROTECTION SYSTEM INSTRGENTATION i

ACTION 9 - In OPERATIONAL COtOITION 1 or 2, be in at least HDr SHtfIDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

In OPERATIONAL CONDITION 3 or 4,

lock the reactor mode switch in the Shutdown position within one hour.

In OPERATIONAL CONDITION 5,

suspend all operations involving 00RE ALTERATIOE or positive reactivity changes and fully insert all insertable control rods within one hour.

TABLE NDPATIONS a.

A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in.Ie same trip system is monitoring that parameter.

b.

'Ibe " shorting links" shall be removed from the RPS circuitry during CORE:

ALTERATIONS and shutdown margin demonstrations performed in accordance with specification 3.10.3.

c.

The IBM scrams are automatically bypassed when the reactor vessel mode switch is in the Run position and all APBM channels are OPERABLE and on scale.

d.

An APPM channel is inoperable if there are less than 2 IPBM inputs per level or less than eleven LPRM inputs to an APBM channel, e.

These functions are not required to be OPERABLE when the reactor pressure vessel head is unbolted or removed.

f.

This function is automatically bypassed when the reactor mode switch is in other than the Run position.

g.

This function is not required to be OPERABLE when PRIMARY CONTAIN4ENT INTH3RITY is not required; this function may be bypassed when necessary for containment inerting or de-inertirg (purging).

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h.

With any control rod withdrawn.

Not applicable to control rods removed per l

Specification 3.9.11.1 or 3.9.11.2.

l

i. These functions are bypassed when turbine first stage pressure is <250* psig, equivalent to WERMAL POWER less than 30% of RATED 'HIERMAL POWER.

j. Also trips reactor coolant system recirculation pump M3 sets.

k.

Also trips reactor coolant system recirculation pump motors.

t 1.

Not applicable while all fuel is removed from the reactor vessel.

l l

Initial setpoint. Final setpoint to be determined during startup testing.

I HA'IGi - UNIT 2 3/4 3-5 AmendmentNo.f,2/,

1

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

TABLE 3.3.2-1 (Continued)

=

ISOLATION AC'IUATION INS'IRIMNPATION VALVE GROUPS MINIEN NLEBER APPLICABLE F

OPERA'IED BY OPERABLE CHANNEIS OPERATIONAL TRIP FUtCTION SIGNAL (a)

PER 'IRIP SYSTD4(b) (c)

CONDITION ACTION 2.

SECONDARY CONTAIWlENT ISOIATION a.

Reactor Buildirg Exhaust l

Radiation - High 6, 10, 12,*

2 1,' 2, 3, 5 and** (j) 24 (2Dll-K609 A, B, C, D) b.

Drywell Pressure - High 2, 5, 6, 7, 10, 2

1,2,3 24 (2C71-N002 A, B, C, D) 11, 12, i,*

c.

Reactor Vessel Water Level - Iow 2, 5, 6, 10, 11, 12,*

2 1,2,3 24 (2B21-N017 A, B, C, D) 6 d.

Refueling Floor Exhaust

'l Radiation - High 6, 10, 12, f,

2 1, 2, 3, 5 and **(j) 24 w4 (2Dll-K611 A, B, C, D) u 3.

REAC'IOR WATIR CLEANUP SYSTD4 ISOLATION a.

Flow - High (2G31-N603 A, B) 5 1

1,2,3 25 b.

Area Tenperature - High 5

1 1,2,3 25 (2G31-N600 A, B, C, D, E, F) c.

Area Ventilation Tenp. - High 5

1 1,2,3 25 (2G31-N602 A, B, C, D, E, F) d.

SLCS Initiation (NA) 5(9)

NA 1,2,3 25 e.

Reactor Vessel Water Ievel - Iow 2, 5, 6, 10, 11, 12 2

1,2,3 25 (2B21-N017 A, B, C, D)

O

HB[E 3.3.2-1 (Continued)

ISO [ATIm ACTtRTION INSTRLNENTATION ACTION ACTION 20 - Be in at least HDT SHMDOti within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in 00[D SHMDOW within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 21 - Be in at least STAlmP with the main steam line isolation valves closed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOP SHWDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COID SHmDolei within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTIm 22 - Be in at least SPARPLP within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

^

ACTION 23 - Be in at least SPARTUP with the Group 1 isolation valves closed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or in at least HOP SimDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

1 i

ACTION 24 - Establish SBCottmRY CONTAIN4ENP INITGRITY with the standby gas treatment system operating within one hour.

ACTION 25 - Isolate the reactor water cleanup system.

ACTION 26 - Close the affected system isolation valves and declare the affected system inoperable.

ACTION 27 - Verify power availability to the bus at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or close the affected system isolation valves and declare the affected system inoperable.

ACTION 28 - Close the shutdown cooling supply and reactor vessel head spray isolation valves unless reactor steam dome pressure 135 psig.

i NOITS Actuates operation of the main control room envirormental control system in the pressurization mode of operation.

Actuates the standby gas treatment system.

- When handling irradiated fuel in the secondary containment.

a.

See Specification 3.6.3.1, Table 3.6.3.1-1 for valves in ead valve group.

b.

A dannel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance without placing the trip system in the tripped condition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter, c.

With a design providing only one dannel per trip system, an inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, the inoperable channel shall be restored to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or the ACTION required by Table 3.3.2-1 for that Trip Function shall be taken.

d.

Trips the mechanical vacuum pumps.

e.

A channel is OPERABLE if 2 of 4 instruments in that dannel are OPERABLE.

f.

May be bypassed with reactor steam pressure 1045 psig and all turbine stop valves closed.

g.

Closes only RWCU outlet isolation valve 2G31-E004.

h.

Alarm only.

i. Adjustable up to 60 minutes.

l

j. Not applicable while all fuel is removed from the reactor vessel.

l HATCH - UNIT 2 3/4 3-15 AmendmentNo./,

TABLE 4.3.2-1 ISO [ATION AC'IUATION INSWINNPATION SURVEILIANCE REQUIRDENTS 4

CHAMEL OPERATIONAL CHAMEL FUNCTIONAL CHAFNEL CONDITIONS IN WHICH E

'IRIP FtRCTION CHECK

'ITST CALIBRATION SURVEILIAICE REQUIRED 4

w 1.

PRIMAIN CONTAIbMENT ISO [ATION 4

a.

Reactor Vessel Water Level i

1. Low D

M O

1,2,3

2. Iow Iow D

M Q

1,2,3 b.

Drywell Pressure - High NA M

Q 1, 2, 3 i

c.

Main Steam Line

1. Radiation - High D

W(a)

R 1,2,3

2. Pressure - Iow NA M

Q l

3. Flow - High D

M Q

1,2,3 d.

Main Steam Line Tunnel

^

Tenperature - High NA R

R 1,2,3 w

h f

t e.

Condenser Vacuum - Iow NA M

Q 1,2,38 f.

Turbine Building Area Tenp. -

High NA M

R 1, 2, 3 2.

SEmtOARY CONTAIMENP ISGATION a.

Reactor Building Exhaust Radiation - High D

M(a)

R 1, 2, 3, 5 and* (b) l j

b.

Drywell Pressure - High NA M

Q 1,2,3 1

l c.

Reactor Vessel Water Ievel -

d Iow D

M Q

1,2,3 d.

Refueling Floor Exhaust Radiation - High D

M(a)

Q 1, 2, 3, 5 and* (b) l

When handling irradiated fuel in the secondary containment.

When reactor steam pressure > 1045 psig and/or any turbine stop valve is open.

(a) Instrument alignment using a standard current source.

(b)Not applicable while all fuel is removed from the reactor vessel.

l

TABLE 3.3.3-1 EMERGEFCY CORE COOLING SYSTEM ACTUATION INS'IRIMNPATION M

MINIKM bENBER APPLICABLE i

OPERABLE OIAl2EIS OPERATIONAL TRIP EUtCTION PER TRIP SYSTEM 00NDITIONS 4

1.

CORE SPRAY SYSTEM w

a.

Reactor Vessel Water Level - Iow Iow Iow (2B21-N031A,B,C,D) 2 1,2,3,4,5 (b) l b.

Drywell Pressure - High (2 Ell-N0ll A,B,C,D) 2 1,2,3

~

c.

Reactor Steam Dome Pressure - Iow (Injection Permissive)

(2B21-N021A,B,C,D) 2 1,2,3,4,5 (b) l d.

Ingic Power Monitor (2E21-KlA,B) 1/ bus (a) 1,2,3,4,5(b) l l

2.

IIM PRESSURE C00DNP INJ1CPION MODE OF RHR SYSTEM a.

Drywell Pressure - High (2 Ell-N0llA,B,C,D) 2 1,2,3 b.

Reactor Vessel Water Ievel - Iow Iow Iow (2B21-NO31A,B,C,D) 2 1,2,3,4*,5* (b) l q

c.

Reactor Vessel Shroud Invel - High (Drywell Spray Permissive)

(2B21-N036 and 2B21-NO37) 1 1,2,3,4*,5* (b) l d.

Reactor Steam Dome Pressure - Iow (Injection Permissive) g 4

(2B21-N021A,B,C,D) 2 1,2,3,4*,5* (b) l cn e.

Reactor Steam Dome Pressure - Iow (Recirc. Discharge Valve Permissive)

(2B21-N021B,C,E,F) 2 1,2,3,4*,5* (b)

I f.

RHR Punp Start - Time Delay Relay 1/punp 1,2,3,4 *,5* (b) l

]

1)

Punp A (2 Ell-K70A, 2 Ell-K125B) j 2)

Punp B (2 Ell-K70B, 2 Ell-K125A) 1 3)

Punp C (2 Ell-K75B) 4)

Punp D (2 Ell-K75A, 2 Ell-K126) j g.

Iogic Power Monitor (2 Ell-KlA,B) 1/ bus (a) 1,2,3,4*,5*(b) l 4

Not applicable when two core spray system subsystems are OPERABLE per Specification 3.5.3.1.

(a) Alarm only. When inoperable, verify power availability to the bus at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or declare the system inoperable.

(b)

Not applicable while all fuel is removed from the reactor vessel.

l l

i

TABLE 4.3.3.-l ENERGEMN OORE COOLING SYSTEM ACIUATION INS'IRWlENTATION SURVEILIMCE REOUIRDDTTS CHANNEL OPERATIONAL

~

8 CHAtNEL EU!CTIONAL CHAMEL CONDITIOtB IN NHICH g

TRIP FUNCTION CHECK TEST CALIBRATION SURVEIILAtCE REQUIRED 1.

CORE SPRAY SYSTEM w

a.

Reactor Vessel Water Level -

Low Iow Low D

M Q

1, 2, 3, 4, 5 (a) l b.

Drywell Pressure - High NA M

Q 1,2,3 c.

Reactor Steam Dome Pressure - Iow NA M

Q 1, 2, 3, 4, 5 (a) l d.

Iogic Power Monitor NA R

NA 1, 2, 3, 4, 5 (a) l 4

2.

IOW PRESSURE COOLANT IMIECTION MODE OF RHR SYSTEM w

a.

Drywell Pressure - High NA M

Q 1,2,3

)

b.

Reactor Vessel Water Level -

Iow Iow Low D

M Q

1, 2, 3, 4*, 5* (a) l Y

c.

Reactor Vessel Shroud Invel -

High D

M Q

1, 2, 3, 4*, 5* (a) l, d.

Reactor Steam Dome Pressure - Iow NA M

Q 1, 2, 3, 4*, 5* (a) l e.

Reactor Steam Dome Pressure - Iow NA M

Q 1, 2, 3, 4*, 5* (a) g f.

RHR Punp Start-Time Delay Relay NA NA R

1, 2, 3, 4*, 5* (a) l g.

Iogic Power Monitor NA R

NA 1, 2, 3, 4*, 5* (a) l

Not applicable when two core spray system subsystems are OPERABLE per Specification 3.5.3.1.

(a) Not @ plicable while all fuel is removed from the reactor vessel.

l i

-m

-n

,__g g_

TABLE 3.3.5-1 00tf1 POL ROD WITHDRAHAL BIIXX INETPRLDENPATION

~

MINIMLM MMBER OF APPLICABLE OPERABIE CHANNEIS OPERATIOtRL TRIP MJfCTION PER TRIP EUTCPION G)NDITIONS Q

l.

APRM (2C51-K605 A, B, C, D, E, F) i w

a.

Flow Referenced Sinulated Thermal 4

1 Power

' Upscale b.

Inoperative 4

1, 2, 5, (g) l c.

Downscale 4

1 d.

Neutron Flux - High, 12%

4 2, 5 (g) l 2.

ROD BLOCK MONI'IOR (2C51-K605 RBM A and B)

J a.

Upscale 1

1(a) w b.

Inoperative 1

1(a) 4 1

c.

Downscale 1

1(a) 3.

SOURCE RANGE MONI'IORS (2C51-K600 A, B, C, D) a.

Detector not full in (b) 3 2

j 2

5(9) l b.

Upscale (c) 3 2

?

2 5(9) l c.

Inoperative (c) 3 2

2 5(9) l d.

Downscale(b) 3 2

1 2

5(9) l 1o 4.

INTERPEDIATE RA!CE MONI'IORS(d) j (2C51-K601 A, B, C, D, E, F, G, H) i a.

Detector not full in (e) 6 2, 5(9) f b.

Upscale 6

2, 5(9)

I c.

Inoperative 6

2, 5 (9) l

[

d.

Downscale 6

2 5.

SCRAM DISCHARGE VOLINE (2Cll-N013E) a.

Water Level-High 1

1, 2, 5(f) (9) l

TABLE 3.3.5-1 (Continued)

CONIROL ROD WITHDRAWAL BIOCK INSIRIMNTATION NOTE a.

When 'IEERMAL POWER exceeds the preset power level of the RM4 and RSCS.

b.

This function is bypassed if detector is reading > 100 cps or the IRM channels are on rarge 3 or higher.

c.

This function is bypassed when the associated IRM channels are on range 8 or higher, d.

A total of 6 IRM instruments must be OPERABLE.

e.

'Ihis function is bypassed when the IBM channels are on range 1.

f.

With any control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.11.1 or 3.9.11.2.

g.

Not @plicable while all fuel is removed from the reactor vessel.

l HA'IGI - UNIT 2 3/4 3-39

TABLE 4.3.5-1

^

5 00tfIROL ROD WI'IMDRKelAL BIOCK INSTRENENTATION SURVEILIANCE REQURIEMENIS CHAM EL OPERATIONAL G

CHANtEL FUICTIONAL CHANNEL CONDITIONS IN WHICH TRIP.FUICTION CHECK TEST CALIBRATION (a)

SURVEILLAICE REQUIRED h APRM a.

Flow Referenced Simulated Thermal Power-Upscale NA S/U (b),M R

1 b,

Inoperative NA S/U (b),M NA 1, 2, 5 (f) l c.

Downscale NA S/U (b),M R

1 d.

Neutron Flux - High, 12%

NA S/U (b),M R

2, 5 2.

ROD DIOCX MONI'IOR 4

a.

Upscale NA S/U (b),M R

1(d) w b.

Inoperative NA S/U (b),M NA 1(d)

D c.

Downscale NA S/U (b),M R

1(d) 3.

SOURCE RAIGE MONI'IORS a.

Detector not full in NA S/0(b),W NA 2, 5(f) l b.

Upscale NA S/U (b),W R

2 5(f) c.

Inoperative NA gjg (b),W NA 2 5(f)

I d.

Downscale NA S/O (b),W R

2, 5(f) l 4.

INTERMEDIATE RAM 3E MONI'IORS 1

a.

Detector not full in NA S/U(b),w(c)

NA 2, 5(f) b.

Upscale NA S/U (b),y(c)

R 2, 5(f)

I c.

Inoperative NA S/U(b),W(c)

NA 2, 5(f) l d.

Downscale NA S/U(b),W(c)

R 2, 5(f) l

[

5.

SCRAM DIL31AIGE VOLLNE 6

a.

Water Level-High NA Q

R 1, 2, 5(e)(f) l 4

I

TABLE 4.3.5-1 (Continued)

CXNIROL RCD WITHDRAWAL BIOCK INS'IRtMNTATION SURVEILIAICE REOUIRIMNTS NOTES:

a.

Neutron detectors may be excluded from CHANNEL CALIBRATION.

b.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to startup, if not performed within the previous 7

days, c.

When changing ~ from COtOITION 1 to CONDITION 2,

perform the required surveillance within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering OJNDITION 2.

d.

When 'IEER!%L POWER exceeds the preset power level of the RM4 and RSCS.

e.

With any control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.11.1 or 3.9.11.2.

f.

Not applicable while all fuel is removed from the reactor vessel.

l l

l I

l l

HA'ICH - UNIT 2 3/4 3-42 1

l l

i

]

s TABLE 3.3.6.2-1 i

SEISMIC MONITORI!G INSTRIMNTATION MINIM M IEASUREMEtTP INSTRGENTS INSTRIMENTS AND SENSOR IOCATIONS RAPGE OPERABLE 1.

Triaxial Time-History Accelerographs (a) a.

Diesel Generator Building El 130'0" (c) 0-0.59 1

(2L51-N021) i b.

Reactor Buil' ding 87' Level on Drywell Pedestal (2L51-N020) 0-0.5g 1 (d) l c.

Drywell - Feedwater Inlet to RPV 0-0.59 1 (d) l (2L51-N004) d.

Switchyard (c)

(lL51-N005) 0-0.59 1

2.

Triaxial Peak Recording Accelermeters a.

Diesel Generator Base Support (c) 0-1.0g 1

(lL51-N007) b.

Intake Structure (c) (lL51-N006) 0-1.0g 1

c.

Control Building Main Control Rom Floor (c) (1L51-N008) 0-1.0g 1

d.

Control Building Floor El 112'(c) 0-1.09 1

(2L51-N028) e.

Reactor Bldg Refueling Floor 0-1.0g 1

(2L51-N029) f.

Reactor Pedestal Inside Biological Shield (2L51-NO35) 0-2.0g 1 (d) l g.

Reactor Piping - Feedwater Inlet to RPV (2L51-N034) 0-2.0g 1 (d) l 3.

Triaxial Seismic Switches (b) a.

Reactor Building 87' Level on Drywell Pedestal (2L51-N022) 0.025-0.25g 1 (d) l b.

Reactor Building 185' Level Outside Biological Shield (2L51-N024) 0.025-0.259 4.

Triaxial Response Spectrum Recorder (a) a.

Hatch - Unit 1 Containment 2-26 Hz 1

Foundation El 87'(c) (lL51-N105) 0-0.59 a With main control room indication and annunciation.

b With main control room annunciation.

c Shared with Hatch - Unit 1.

Not applicable while all fuel is removed from the reactor vessel.

l d

HATOI - UNIT 2 3/4 3-48 Amendment No.

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

i s

TABLE 3.3.6.8-1 FIRE DETECTION INS'IREMNTATION MINIMM OPERABLE AREA NtNBER OF DETECTORS Dr; m ;1unS Control Building, El. 112'-0" Station Battery Room 2A 2

1 Station Battery Room 2B 2

1 Corridor and Work Area 9

4 Control Building, El. 130'-0" Switchgear Roms, 2A, 2B, 5

1 per room 2C, 2D and Transformer Room Corridor 6

3 RPS Vertical Cableway 2

1 East Cableway 14 7

Diesel Generator Building Switchgear Rom 2E 4

2 Switchgear Room 2F 4

2 Switchgear Rom 2G 4

2 Battery Room 2A 1

1 Battery Rom 2C 1

1 Reactor Building, El. 130'-0" North Cable Tray Arca, Remote Shutdown Area, and North CRD Area 38 19 South Cable Tray Area and South CRD Area 35 17 Reactor Building, El. 156'-0" HVAC Room 23 11 i

l Drywell Recirculation Pung A Area 1

1 (a) l Recirculation Pump B Area l

1 (a) l Div. 1 Control & Power Cable Tray Area 1

1 (a) l Div. 2 Control & Power Cable Tray Area 1

1 (a) g Div. 1 CRD Cable Tray Area 1

1 (a) l Div. 2 CRD Cable Tray Area 1

1 (a) l (a) Not applicable while all fuel is removed from the reactor vessel.

l HA101 - UNIT 2 3/4 3-60 Amendment No.

!'O

s DERGENCY CORE COOLING SYSTEMS IfW PRESSURE COOIANT INJECfION SYSIH4 LIMITING COFDITION IOR OPERATION 3.5.3.2 Wo independent Low Pressure Coolant Injection (IKI) subsystems of the residual heat removal system (RHR) shall be OPERABLE with each subsystem conprised of:

a.

W o OPERABLE RHR pumps, b.

An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor pressure vessel.

APPLICABILITY: CONDITIONS 1, 2, 3, 4* and 5*, ** (a),

l ACTION:

a.

In CONDITION 1, 2, or 3; 1.

With one LPCI subsystem or one LPCI punp inoperable, POWER OPEPATION may continue provided both CSS subsystems are OPERABLE; restore the inoperable LPCI subsystem or punp to OPERABLE status within 7 days or be in at least HOT SHMDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

With both LPCI subsystems inoperable, be in at least HOf SIUIDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and either be in COID SHUIDOWN or maintain reactor coolant tenperature f,4000F by use of alternate heat removal methods within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.

With the LPCI system cross-tie valve open or power not removed from the valve operator, be in at least HOT SHUIDOWN within 12

[

hours and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l l

4.

In the event the LPCI system is actuated and injects water into l

the Reactor Coolant System, a Special Report shall be prepared and subnitted to the Comnission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

b.

In CONDITION 4*

or 5*,

- with one or more LPCI subsystems inoperable, take the ACTION required by Specification 3.5.3.1.

The provisions of Specification 3.0.3 are not applicable.

Not applicable when two CSS subsystems are OPERABLE per Specification i

l 3.5.3.1.

- Not applicable when the CSS is not required to be OPERABLE per Specification 3.5.3.1.

l (a)Not applicable while all fuel is removed from the reactor vessel.

l t

l HA'IQI - UNIT 2 3/4 5-7 I

i

i s

REFUELIE OPERATIOE 3/4.9.2 INSTRINENTATION LIMITING COmITION FOR OPERATION 3.9.2 At least 2 source range monitor * (SRM) channels shall be OPER@LE and inserted to the normal operating level:

a.

Each with continuous visual indication in the control room, b.

At least one with an audible alarm in the control room, c.

One of the SRM detectors located in the quadrant where CDRE ALTERATIOE are being performed and the other SRM detector located in an adjacent quadrant, and d.

The "shortirg links" removed from the RPS circuity during CORE

~

ALTERATIO E and shutdown margin demonstrations.

APPLICABILIq: COEITION 5(a),

l ACTION:

With the requirements of the above specification not satisfied, imediately suspend all operations involving CORE ALTERATIONS ** or positive reactivity charges and actuate the manual scram.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLAbCE FwUINEMENTS 4.9.2 Each of the above required SRM channels shall be demonstrated OPERABLE by:

a.

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; 1.

Performance of a CHANNEL CHECK, 2.

Verifying the detectors are inserted to the normal operating

level, 3.

During CORE ALTERATIOE, verifying that the detector of an OPERABLE SRM channel is located in the core quadrant where CORE ALTERATIOE are being performed and one is located in the adjacent quadrant.

The use of special movable detectors during CORE ALTERATIONS in place of the normal SRM nuclear detectors is permissible as long as these special detectors are connected to the normal SRM circuits.

- Except movement of SRM or special movable detectors.

(a) Not applicable while all fuel is removed from the reactor vessel.

l HATUI - UNIT 2 3/4 9-3

REFUELING OPERATIONS 3/4.9.3 CONTR % IOD POSITION LIMITING CONDITION FOR OPERATION 3.9.3 All control rods shall be fully inserted.*

APPLICABILITY: CONDITION 5, during CORE ALTERATIOtE.**(a) l ACTION:

With all control rods not fully inserted, suspend CORE ALTERATIONS.

The provisions of specification 3.0.3 are not applicable.

SURVEILIRCE t<EUUHREMENTS 4.9.3 All control rods shall be verified to be fully inserted within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the start of and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during COPE ALTERATIONS.

i l

Except control rods removed per Specification 3.9.11.1 or 3.9.11.2.

- See Special Test Exception 3.10.3.

(a)Not applicable while all fuel is removed from the reactor vessel.

l f

HATUI - UNIT 2 3/4 9-5

t REFUELING OPERATIOtB 3/4.9.12 REAC'IOR 000DWT CIRCULATION LIMITIfG COBOITION FOR OPERATION 3.9.12 The residual heat removal (RHR) system shall be OPERABLE with at least one OPERABLE punp and one OPERABLE heat exchanger.

l APPLICABILITY: OPERATIONAL CONDITION 5*.

ACTION:

a.

With the residual heat removal system inoperable, suspend all operations involving an increase in the reactor decay heat load or a positive reactivity change.

Close all secondary containment penetrations providing direct access from the secondary containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLA?CE REOUIREMENTS 4.9.12

'Ibe residual heat removal system shall be determined OPERABLE by verifying at least once per 31 days that:

a.

At least one RflR system punp can be started from the control room, if not already operating, and b.

System valves are properly aligned to provide recirculation of reactor coolant through the RHR heat exchanger.

Not applicable while all fuel is removed from the reactor vessel.

l HA'ICH - UNIT 2 3/4 9-20

5 ENCIOSURE 2 NRC DOCKET 50-366 OPERATING LICENSE NPF-5 EDNIN I. HA'ICH NUCLEAR PLANT UNIT 2 REQUEST 'IO AMEND TECHNICAL SPECIFICATIONS Pursuant to 10 CFR 50.92, Georgia Power Cmpany has evaluated the attached proposed amendment and has determined that its adoption would not involve a significant hazard.

The basis for this determination is as follows:

PROPOSED CHANGES Change the applicability of the limiting conditions for operation of systes and instrumentation associated with the cooling and protection of fuel in the reactor vessel after all fuel is reoved frm the vessel.

BASIS Since all fuel will be rmoved frm the reactor vessel and stored in the fuel pool during the main recirculation syste piping replacment scheduled tentatively for January 1984 and at other times, as appropriate, equipnent such as RHR, Nuclear Instrmentation, Scra Systs, Standby Liquid Control Systs, Control Rod Drive Systs, and ECCS Actuation are not needed to prevent a fuel-related accident while all fuel is reoved frm the vessel.

Deletion of the limiting conditions for operation of the aformentioned systems and instrumentation while the fuel is reoved frm the reactor vessel does not affect the probability or consequences of an accident or malfunction analyzed in the FSAR.

These changes do not create the possibility of an accident or malfunction of a different type than any analyzed in the FSAR. The margin of safety as defined in the basis for any Technical Specification is not affected.

The effect of these changes is therefore within the acceptance criteria and the changes are consistent with Its (vi) of the "Exmples of Amendments That Are Not Likely to Involve Significant Hazards Considerations" listed on page 14870 of the April 6, 1983 issue of the Federal Register.

4 s

ENCIDSURE 3 NIC DOCKET 50-366 OPERATING LICENSE NPF-5 EDKIN I. HA'ICH NUCLEAR PIANT UNIT 2 REQUEST 'IO AMEND TECHNICAL SPECIFICATIONS Pursuant to 10 CFR 170.22, Georgia Power Cmpany has evaluated the attached proposed men &ent to Operating License NPF-5 and has determined that:

The proposed men &ent does not rquire evaluation of a new Safety a.

Analysis Report or a rewrite of the facility license; b.

He proposed men &ent. does not rquire evaluation of several cmplex issues, involve ACRS review, or require an environmental impact statement; ne proposed men &ent involves one safety issue, nmely changing c.

the Technical Specifications applicability of the limiting conditions for operation of systes and instrunentation associated with the cooling and protection of fuel in the reactor vessel after all fuel is reoved fra the vessel.

d.

%e proposed men &ent is therefore a Class III men &ent.

v t e Letter Sequence Other
TAC:52483, (Open)
Initiation Request, Request, Request, Request, Request Acceptance... Administration Meeting, Meeting Questions Answers Results Approval... Other: ML20080Q819, ML20081B157, ML20081K549, ML20083C544, ML20083C551, ML20083H377, ML20129E491
MONTHYEAR1983-11-07 [Table View]

ML20081B157

Text

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