ML20045E672

Amend 175 to License DPR-50,extending Interval for Performance of Selected Surveillances to Coincide w/24-month Operating Cycle,Per GL 92-04

ML20045E672
Person / Time
Site:	Crane
Issue date:	06/23/1993
From:	Stolz J Office of Nuclear Reactor Regulation
To:	Metropolitan Edison Co, Jersey Central Power & Light Co, Pennsylvania Electric Co, GPU Nuclear Corp
Shared Package
ML20045E673	List: ML20045E672 ML20045E683
References
GL-92-004, DPR-50-A-175 NUDOCS 9307020300
Download: ML20045E672 (21)

Text

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'E UNITED STATES W

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?i NUCLEAR REGULATORY COMMISSION

- WASHINGTON, D.C. 20555-0001 e.....;j METROPOLITAN EDISON COMPANY.

' JERSEY CENTRAL POWER & LIGHT COMPANY.

..g PENNSYLVANIA ELECTRIC COMPANY GPU NUCLEAR CORPORATION DOCKET NO. 50-289 THREE MILE ISLAND NUCLEAR STATION. UNIT NO. I g

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AMENDMENT TO FACILITY-OPERATING' LICENSE I.

. Amendment No.175:

License.No..DPRtSOL 4t 1.

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

A.

'The application for. amendment' by.GPU Nut. lear Corporation, eMali,2

-(the'1icensee)' dated: June'24,z1993, as supplemented by.l.etteri dated May 28,1993,1 complies with the standards and Lrequirements li of the Atomic: Energy Act of 1954,.as! amended-(the;Act). and the L

. Commission'~s rulessand regulations set forth inL10 CFR Chapter I.;-

n B.

The facility will operate in co6formity with':the application, ther provisions of the Act, and the. rules.and.. regulations. off the' Commission;:

C.

Thereisr'easonableassurance(i)'.thititheactivitiesauthorizedi

~

by this amendment can be conducted:without endangeri_ng the health' and safety of the:public, and (ii)lthat such activities willt be:

-conductedincompliance.withtheCommission'sregulations;j D.

The, issuance:of.this. amendment will: not be : inimical to; the: common?

defense!and security;or toLthe health and safe _tyfof thelpublic;-

and:-

p E,

.The issuance-of this~ amendment is.;jn':accordance.with 10 CFR Part 51 of-the. commission's : regulations'and.-all c pplicable requirementsa a

,s have beenLsatisfied.

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,p ADOCK 05000289:

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2.

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 Operating License No.

DPR-50'is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A,.as revised through Amendment No.175, are hereby incorporated in the license.

GPU Nuclear Corporation shall operate the facility in accordance with the Technical Specifications.

P 3.

This license amendment is effective as of its date of issuance, to be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION n'F. Stolz, Directo Jo/ojectDirectorateI-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: June 23, 1993

?

.m A

ATTACHMENT TO LICENSE AMENDMENT NO. 175-

-FACILITY OPERATING LICENSE'NO.-DPR-50 DOCKET NO. 50-289

. Replace the following pages of the Appendix A Technical Specifications-with

~

the attached.page. The revised pages are identified.by.an amendment number'-

1

. and contains vertical lines indicating the area.of. change.

q E9.!! Ley.ft Insert-1-P 1 *

'l-8 8 1-9 l' 9 -

4-3 4-3 4-4

.4-4 4 4-5 4-Sa 5a:

4-6 4-6 4-7' 4-7 4-7a:

4-7a-4 4-8 4-10a' 10a 4-10b 4-38 4-38 4-47 4 4-55 4-55' 4-55b

'4-55b 4-60

~4 '

3 0

e t

i.2.'

1 1

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L

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+

, L.

1, 4. :-

1.2.7-REFUELING OPERATION An. operation' involving a. change in core geometry by 6anipulation of fuel or-control rods.when the reactor vessel head :is' removed, l'.2.8 REFUELING-INTERVAL Time between normal refuelings of the reactor. This is defined as once 'per 24 months.

1.2.9 STARTUP l

The reactor.'shall be considered in the startup mode when the shutdown margin'.

is reduced with the intent of going critical.

l 1.2.10 T,y, T,, is defined as the arithmetic average of the coolant temperatures in the!

hot and cold legs of the loop with the greater number of reactor coolant-pumps operating, if such a distinction of loops can be made.

l.2.11 HEATUP - C00LDOWN MODE The heatup-cooldown mode is the range of reactor coolant temperature.

greater-than 200*F and less than 525'F.

1.2.12 STATION, UNIT, PLANT,.AND FACILITY f.

Station, unit, plano, and _ facility as used in these technical specifications all refer to TMI Unit 1.

1.3 OPERABLE

. A system, subsystem, train, component or device shall be OPERABLE or[have~

OPERABILITY when it is capable of performing.its specified function (s) Land.-

when al.1 necessary attendant instrumentation,.. controls, electrical. power,-

cooling or seal water, lubrication.or other ' auxiliary equipment that are o

required-for the system, subsystem, train, component, or_ device to perform its function (s) are also. capable. of. performing their related: support function (s).

-1.4 PROTECTION INSTRUMENTATION LOGIC.

1.4.1 INSTRUMENT CHANNEL An' instruments channel is_ the combination-of. sensor, wires, amplifiers,' and;..

output devices'whichc are connected for' the purpose of measuring the..value, of-a processLvariable for the purpose of observation, control, and/or:

protection. -' An instr'ument channel may be either analog or digital.

_a 1

1-2

AmendmentNo.9$,if/,'175

+

4

1.24

. CORE OPERATING LIMITS REPORT The CORE OPERATING ~ LIMITS REPORT is a TMI-1 specific document that provides-core operating limits for the current operating reload cycle.

These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.5.

Plant operation within these operating limits is addressed in individual specifications.

1.25 FREQUENCY NOTATION The FREQUENCY NOTATION specified for the performance of Surveillance kequirements shall correspond to the intervals defined.in Table 1.2.

All Surveillance Requirements shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the surveillance interval. The 25% extension applies.to all frequency intervais with the exception of "F."

No extension is allowed for intervals designated "F."

TABLE 1.2 FRE0VENCY NOTATION t

NOTATION FREQUENCY S

Shiftly (once~per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />)

D Daily (once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)

W Weekly (once per 7 days)

M Monthly (once per 31 days)

Q Quarterly (once per 92 days)

S/A Semi-Annually (once per 184 days)

R Refueling Interval (once per 24 months)

P S/U Prior to each reactor startup, if not-done during the previous 7 days P

Completed prior to each release N/A (NA)

Not' applicable E

Once per 18 months F

Not to exceed 24 months 1-8 Amendment No. 77, 177, 755, 177, 175

Bases-Section 1.25 establishes the limit for which the specified time interval for Surveillance Requirements may be extended.

It permits an allowable extension of the normal surveillance interval to facilitate surveillance scheduling and consideration of plant operating conditions that may not be suitable for conducting the surveillance; e.g., transient conditions or other ongoing surveillance or maintenance. activities.

It also provides flexibility to accommodate the length of a fuel cycle for surveillances that are specified to be performed at least once each REFUELING INTERVAL.

It is not intended that this provision be used repeatedlyfas a convenience to extend surveillance intervals beyond that specified for surveillances that are not performed once each REFUELING INTERVAL. Likewise, it is not the i

intent that REFUELING INTERVAL surveillances be performed during power operation unless it is consistent with safe plant operation.

The limitation of Section 1.25 is based on engineering judgement'and the recognition that the most probable result of any particular surveillance being performed is the verification of conformance with the Surveillance Requirements.

This provision is sufficient to ensure that the reliability ensured through surveillance activities is not significantly degraded beyond that obtained from the specified surveillance interval.

P 4

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I i

r 1-9 Amendment No. /f, ff/, Iff, (/f, 175 nr

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TABLE-4.1-1 o

S INSTRUMENT SURVEILLANCE REQUIREMENTS A

f CHANNEL DESCRIPTION CHECK TEST CALIBRATE REMARKS 1.

Protection Channel NA' M

NA

'y Coincidence logic PJ 2.

Control Rod Drive Trip NA N

NA (1)

Includes independent testing 'of shunt Breaker and Regulating trip and undervoltage trip features.

y Rod Power SCRs 3.

Power Range. Amplifier D(1)

NA (2)

(1) When reactor power is greater than 15%.

~

(2) When above 15% reactor power run a heat U

balance check once per shift. Heat balance calibration shall be performed-whenever heat balance exceeds indicated neutron power by more than two percent.

[

4.

Power Range Channel S

M M(1)(2) (1) When reactor power is greater than 60%

verify imbalance using incore

. instrumentation.

(2) When above 15% reactor power calculate

axial offset upper and lower chambers after each startup if not done within the previous seven days.

5.

Intermediate Range Channel S(l)

PS/U NA (1). When in' service.

6.

Source Range Channel S(l)

PS/U NA

.(1) When in service.

7.

Reactor Coolant Temperature S

M F

Channel r

~

\\

j.

t

' 1 TABLE'4.1-1 (Continued) 3 :s ;k CHANNEL DESCRIPTION'

- CHECK TEST-CALIBRATE REMARKS.

x S

M.

R 8.

High. Reactor Coolant 2o-Pressure Channel S

9.

Low Reactor: Coolant S-M-

R

-w Pressure Channel m.

L L10. Flux-ReactorT oolant' Flow.

S M

F C

. 7-

'Comparator w:

.g

,.11. (Deleted).

h

12. Pump Flux Comparator.

- S fM-R

.w

13. High Reactor.' Building'.

S M'

F Pressure Channel

14._HighlReactor' Building NA' Q

NA f*

Logic Channels-

~

l 2

3

-15.iHigh' Pressure Injectioni Analog" Channels a.

Reactor Coolant.

S(l)

M R-(1) When reactor coolant-system is -

Pressure Channel" pressurized-above 300 psig or T,,is.

greater than 200*F.

16. Low Pressure Injection

- NA'.'

-Q.

NA.

Logic Channel"

^

1 zl7. Lower. Pressure Injection.

Analog:Channelsi ia. l Reactor / Coolant'

. S(1)>

M

- R-(1) When' reactor coolant ~ system is I. Pressure-Channel

~

pressurized above 300 psigior T is:

m greater:than:200*F..

-:18., Re' actor; Building, Emergency;-

. NA:.

,Q l NA ~:

iCooling~'and Isolation System-4 Logic Channel.

~

.I f.C '..

.c

• ..v-

...e..-

- - * -..,. -.. ~

e--a

+*+ -=

r

-e-+-

a +~ -

" + -

+-

i

++r

~

L.

TABLE 4.1-1 (Continued);

~g CHANNEL DESCRIPTION-CHECK IEST-CALIBRATE REMARKS h

19. Reactor Building 1 Emergency 2L Cooling and Isolation.

[

System Analog Channels

a. Reactor Building S(l)

M(1)

F.

(1) When CONTAINMENT INTEGRITY is required.

4'psig Channels

.b. RCS Pressure 1600 psig S(l)

M(1)

NA (1) When RCS Pressure > 1800 psig.

M

c. RPS Trip S(1)

M(1)

NA (1) When CONTAINMENT. INTEGRITY'is required.

~

d. Reactor Bldg. 30 psig S(l)

M(1)

F (1) When CONTAINMENT' INTEGRITY is required.

O

e. Reactor Bldg. Purge W(1)

M(1)

F (1) When CONTAINMENT INTEGRITY is required.

Line High Radiation (AH-V-1A/D)

~

f. Line Break Isolation W(1)

M(1)

R (1) When-CONTAINMENT INTEGRITY is required.

~

Signal' (ICCW & NSCCW) 20.' Reactor building Spray.

NA Q

NA g

System logic Channel

21. Reactor Building Spray System Analog Channels f

a. Reactor Building NA M

F 30 psig Channels

22. Pressurizer Temperature S

NA R

-Channels

23. Control' Rod-Absolute Position S(l)

NA R

(1) Check with Relative Position Indicator.

24. Control: Rod Relative Position S(l)'

NA R

(1) Check with Absolute Position Indicator.

25. Core Flooding Tanks a.; Pressure Channels S(1)

NA F

(1) When Reactor Coolant system pressure is greater than 700 psig.

'b. Level Channels S(l)'

NA

.F

26. Pressurizer Level Channels-S NA R

Y TABLE 4.1-1 (Continued)

.3

?

CHANNEL DESCRIPTION CHECK TEST CAllBRATE REMARKS

27. Makeup Tank Level Channels D(1)

NA F

(1) When Makeup and Purification Systcm is l

n operation.

2

28. Radiation Monitoring Systems

• W(1)(3) M(3)

Q(2)

(1) Using +he installed check source when background is less than twice the expected increase in cpm which would

[

result from the check source alone.

Background readings greater than this

?*

value'are sufficient in themselves to E

show that the monitor is functioning.

~

~

5 (2)

Except area gamma radiation monitors RM-G5, RM-G6, RM-G7 and RM-G21 which are located in the Reactor Building. When purging is permitted per T.S. 3.6, RM-GS

~O and RM-G21 will be calibrated quarterly.

a If purging is not permitted per T.S.

% 0, 3.6, RM-G5 and RM-G21 shall be

~

~.

calibrated at the next scheduled reactor G

shutdown following the cuarter in which calibration would normally be due.

RM-G6 and RM-G7, which are in high radiation areas shall be calibrated at the next scheduled reactor shutdown following the quarter in which calibration is due, if 'a shutdown during the quarter does not occur.

(3)

Surveillances are required to be performed only when containment integrity is required. This applies to monitors which initiate containment isolation only.

29. High and Low pressure N/A N/A F

Injection. Systems:

Flow Channels-

• Does not include the monitors covered under Specification 3.5.5.2 and 4.1.3 or Specification 3.21.1, 3.21.2 and 4.21.1, 4.21.2.

<;~.

m-n.

,yy m

m

?-l S

~s TABLE 4.1-1 (Continued)

CHANNEL DESCRIPTION CHECK TEST CAllBRATE REMARKS

?j i30. Borated Water Storage W

NA F

3-Tank Level, Indicator-Y

'31. Boric Acid ~ Mix Tank.

[.

a. ' Level Channel

. NA NA F-b'. Temperature Channel-

~ M 1NA F-r32J Reclaimed Boric Acid

Storage' Tank:

.a. Level Channel

NA NA F

b. Temperature Channel M

NA' F

y

33. Containment; Temperature NA-

NA F

'34. Incore Neutron Detectors M(1)_

~ NA NA' (1) Check functioning; including functioning:

~

of computer readout or recorder readout when reactor.. power.is-grater.than.15%.

35., Emergency Plant ~ Radiation M(1)-

NA-F

' (1) Battery check.

Instrxnents-

36. Strong Motion Accelerometer' Q(1)

.NA~

Q

(1) Battery check.

?37.-. Reactor Bui* Jing Sump-NA-

.NA

-R Level ~.

~

.v.

• =

1 wh-

• 1 C

~ ' ' ' " " " " ' '

• * ' " * ^ * ' ' " "' "

g.

r w

5 1

TABLE:4.1-1 (Continued)

g.

1

- CHANNEL ~ DESCRIPTION CHECK TEST CALIBRATE REMARKS X

38.'OTSG Full-~ Range Level-W

.NA

'R F

39 Turbine.0verspeed Trip NA R

NA

40. ; BWST/NaOH - Differential NA NA F

.M Pressure Indicator 141'. Sodium Hydroxide Tank NA

-NA F

Level-Indicator

42. Diesel Generator Protective NA NA R

g.

Relaying

-w.

~~

43. 4 KV ES. Bus Undervoltage

. Relays (Diesel Start) j

'p

- a. Degraded Grid.

NA M(1)

R' (1)

Relay operation will be-checked by local.

test pusnbuttons.

u

b. Lossc of-Voltage '

NA M(1)

R (1) Relay. operation will be checked by local' test pushbuttons'.

. 44. Reactor Coolant Pressure

S(l)J M

R (1) -When reactor coolant system is DH Valve Interlock Bistable pressurized above 300 psig or T,is

. greater - than.:200*F.

45. Loss of:Feedwater Reactor' Trip

'S(l)

.M(1)

R (1) When reactor power exceeds 7%. power.

46. Turbine Trip / Reactor Trip:

lS(l)-

-; M(1)_

F

- (1) When_ reactor. power exceedsL45% power.

~

- 47.-a. Pressurizer Code Safety Valve S(l)'

NA F

' (1) When T,g is greater than 525*F.

and.-PORY Tailpipe Flow Monitors b;-PORV.- Acoustic / Flow

' NA -

M(1)-

R;

- (l); When-T is-greater than 525'F.-

48.~PORV,Setpoints-NA' M(1) -.

R (1). Per Specification 3.1;12 excluding valve; i

toperation.;-

i't--'g y

e-

+Y'--

w--e v e e

e e-p t's--

e s

tt-re-gr P-

• t'hrr-p

'T W

• -iw--S iw--'Ww

p-

. TABLE 4.1-1 (Continued)

CHANNEL DESCRIPTION CHECK' TEST CALIBRATE

-REMARKS 7*

49. Saturation Margin. Monitor S(l)

M(1)

R' (1) When T,,, is greater than 525aF.

50. Emergency Feedwater Flow NA

.M(1)

F (1) When T,,,is greater than 250*F.

g Instrumentation g

51. Heat: Sink Protection System

?

a. EFW Auto' Initiation (1)

Includes logic test only.

~

M Instrument Channels-

1. Loss of Both Feedwater Pumps NA Q(1)

F

{

0

2. Loss of All RC Pumps NA Q(1)

R 7

3. Reactor Building Pressure NA Q

F

[

4. OTSG Low Level W

_ Q R

wM P

b. MFW Isolation OTSG Low Pressure NA Q

R c.'EFW Control Valve Control System

1. OTSG Level. loops

.g

2. Controllers

. W Q

R W

NA R

d. HSPS Train Actuation Logic NA Q(1)

R

52. Backup Incore Thermocouple Display :M(1)

NA R

'(1) When Ts,is greater than 250*F.

53. Chlorine Detection System W

M

.F(l)

(1) Calibration is a one concentration point Instrumentation check (need not' be' traceable.to NBS-standards).

54. RCS Inventory Trending System

.a. Level NA

-NA F.

b. Void Fraction.

W NA

.F

4 pv TABLE'4.1-2~

t"<'

MINIMUM EQUIPMENT TEST FREQUENCY

[

7.c Jigm T 31 Freauency o

3 1.

Control. Rods Rod' drop times of all

.Each Refueling: shutdown' full length rods.

1 o

2.

Control-Rod Movementzof each rod 1 Every two weeks, when!

Movement' reactor is critical.

3.

Pressurizer Setpoint*

~50% Leach refuel _ing period.'

Safety Valves-4.

Main Steam Setpoint

'Approximat'ely 50% each:

4 Safety Valves refueling; period 5.

Refueling System. Functional Start'ofieach 4

Interlocks refueling period-6.

Main Steam (See.Section 4.8)

Isolation Valves 7.

Reactor Coolant ~

Evaluate

. Daily,lwhen reactor-System Leakage coolant system-Jtemperature

.is; greater'than:525'F;

. B.

(Deleted)

- 9.

Spent' Fuel Functional

'Each refueling period:

Cooling System

- prior.tol fuel handling

10. Intake Pump.

.(a) SiltiAccumulation-Not-tolexceed 24 month's:

I

' House FloorL Visual' inspection (Elevation -

of-Intake. Pump:

~3

'262:ft.-6 in.).

ihouse. Floor -

-(b)- Silt Accumulation'. Quarterly-I

' Measurement 1of

Pump' House flow i

11. Pressurizer Block Functional **

Quarterly.

Valve -(RC-V2)

~

• The setpoint-of the pressurizer. code safety? valves shall be in-i

-accordance with ASME-Boiler and Pressurizer: Vessel: Code, Section 1

III,LArticle 9,; Winter, 1968..

• • Function shall. beidemonstrated by' operating; the valve through.

one complete cycle of= full: travel.

I 4-8 Amendment. No. 55,i R,.(]p,0 f4'9,1753 i

4

.c 34 s

[

TABLE'4 1-4:

.3

[

. POST ACCIDENT-MONITORING-INSTRUMENTATION

^

E.

5,5.

FUNCTION.

INSTRUMENTS CHECK TEST CAllBRATE REMARKS-1.

Noble Gas Effluent

-1 a.

Condenser Vacuum Pump Exhaust W

M F

(1) Using theLinstalled c' heck

=[:<?

(RM-AS-Hi) source when backgroundLis..less:

thari twice the: expected increase in cpm which would result from the check source alone.

Background readings greater than-

.this value :are sufficient -in -

'themselves to show that<this'-

I.

Condenser Vacuum Pump-Exhaust

- W(1)-

M F

b

-(RM-G25) g

c. ~ Auxiliary.and Fuel ~ Handling; W.

M F

s Building Exhaust'(RM-A8-Hi)-

1

' o d.

Reactor. Building Purge Exhaust

.W'

.M F

.(RM-A9-Hi)-

Reactor Building l Purge: Exhaust.

W(1)-

L M --

F e.

(RM-G24)

f..: Main Steam Lines Radiation

.W(1)

'M'-

F.

(RM-G26/RM-G27) 2-

. Containment High Range Radiation W-M R-(RM-G22/G23)-

-3.

Cont'ainment Pressure Wj

'N/A F

l 4.-

' Containment. Water: Level -

- W t-N/A R-

!~

5.-

. Containment 1 Hydrogen W

M F-

~6.

Wide' Range _ Neutron Flux W.

N/AL F.

.....m 2

u

F-

['[ _'

TABLE 4ll-4 (Continued)

~

POST ACCIDENT MONITORING INSTRUMENTATION R

5F FUNCTION INSTRUMENTS CHECK TEST CALIBRATE REMARKS jd) 7.

? Reactor CoolantLSystem Cold Leg W

N/A R

Water Temperature

~5 (it-959,961;-TI-959A,~961A)

.d?-

~

18.

Reactor Coolant System Hog-Leg W-N/A R

g[

(TE-958, 960; TI-958A, 960A)

. 9.

Reactor Coolant System Pressure W

N/A R

'(PT-949, 963; PI-949A, 963) 10.

Steam Generator 1 Pressure 11 N/A R

(PT-950, 951,.1180, 1184; PI-950A,--951A, 1180,-1184)'

j[

11.

Condensate Storage Tank Water. Level W

N/A F

g?-

(LT-1060,J1061. 1062, 1063;-

LI-1050,;1061, 1062, 1063) 4 e

e 7

v y-p8-

'-e y

+y g-

+

esn-y e'

y

+y f

y w

w

%-,yw

'4. 4. 4-Hydrooen Recombiner System Aeolicability Applies to the testing of the hydrogen recombiner and associated controls.

Ob.iective To verify that the hydrogen recombiner and associated controls.

are operable.

4.4.4.1 Soecification a.

At least once per 6 months,-perform a hydrogen recombiner system functional test to demonstrate that the minimum.

reaction chamber gas temperature is maintained 1 600*F for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

b.

At least once per refueling interval, perform the following surveillances:

1.

A channel calibration of all recombiner instrumentation and control. circuits' (interval not to exceed 24 months).

2.

Verify through a visual examination that there is no evidence of. abnormal conditions (i.e., loose wiring or:

structural connections, deposits of foreign materials, etc.)

3.. Verify during a recombiner system functional test that' the reaction chamber gas temperature is maintained 11200*F for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

4.

Verify the integrity of the-heater electrical circuits by performing a continuity and resistance to ground test.

The resistance to ground for any heater phase shall be 110,000 ohms.

E1H1 The surveillance program described above provides high assurance that.

the hydrogen recombiner system will be available to perform its post-LOCA function of maintaining the containment hydrogen concentration below 4.1 volume percent. This system is not credited to mitigate any accident analyzed in Chapter 14. of theLTMI-1 FSAR. The frequency of the surveillance of the-hydrogen'recombiner system is-based on the safety significance of the system. TMI-1.FSAR Section 6.5.3.1 indicates that the hydrogen recombiner system is not required until 9.8. days following a LOCA.

This is adequate time to place a hydrogen recombiner~in service.

4-38 Amendment No. f7, I6s; 175

d.

The battery will be subjected to a load test on a refueling interval basis.

(1) Verify battery capacity exceeds that required to meet design loads.

(2) Any battery which is demonstrated to have less than 85% of manufacturers ratings during a capacity discharge test shall be replaced during the subsequent refueling outage.

4.6.3 Pressurizer Heaters a.

The following tests shall be conducted at least once each refueling:

(1)

Pressurizer heater groups 8 and 9 shall be transferred from the normal power bus to the emergency power bus and energized. Upon completion of this test, the heaters shall be returned to their normal power bus.

(2) Demonstrate that the pressurizer heaters breaker on the emergency bus cannot be closed until the safeguards signal is bypassed and can be closed following bypass.

(3)

Verify that following input of the Engineered Safeguards Signal, the circuit breakers, supplying power to the manually transferred loads for pressurizer heater groups 8.and 9, have been tripped.

Bases The tests specified are designed to demonstrate that one diesel generator will provide power for operation of safeguards equipment. They also assure that the emergency generator control system and the control systems for the safeguards equipment will function automatically in the event of a loss of normal a-c station service power or upon the receipt of an engineered safeguards Actuation Signal. The automatic tripping of manually transferred loads, on an. Engineered-Safeguards Actuation Signal, protects the diesel generators from a potential overload condition. The testing frequency specified is intended to identify and permit correction of any mechanical or electrical deficiency before it can result in a system failure. The fuel oil supply, starting circuits, and controls are continuously monitored and any faults are alarmed and indicated.

An abnormal condition in these systems would be signaled without having to place the diesel generators on test.'

Precipitous failure of the station battery is~ extremely unlikely. The surveillance specified is that which has been demonstrated over the years to provide an indication of a cell becoming unserviceable long before it. fails.

The PORV has a remotely operated block valve to provide a positive shutoff capability should the relief valve become inoperable. The electrical power for -

both the relief valve and the block valve is supplied from an ESF power source to ensure the~ ability to seal this possible RCS leakage path.

The requirement that a minimum of 107 kw of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation.

4-47 Amendment No. 7S, J57,157,175

g

-4.12 AIR TREATMENT SYSTEM-s 4.12.~ 1 EMERGENCY! CON 1ROL ROOM AIR TREATMENT-SYSTEM-

=!

Aeolicability Applies t'o the emergency control room air treatment system ~and associated

~

components.

.I Ob.iective i

To verify that this system and associated components will-be able to perform.

its design functions.

i

-'Soecification i

4.12.1.1 At'least every refueling interval, the pressure drop across.

the ccmbined HEPA-filters and charcoal adsorber banks of -

AH-F3A and '3B shall be demonstrated to'be less than 6 inches -

P of water at.~ system design flow rate ( 10%).

4.12.1.2 a.

The tests and sample analysis 1 required by Specification 3.15.1.2 shall be performed initially and at least once

~

per year for. standby service or' after everyJ720 hours 'of.

system operation and following.-significant painting; steam, fire or chemical. release in?any ventilation zone.

communicating with the system that'.could contaminate:the; HEPA filters or charcoal adsorbers.

b.

00P testing shall be performed after.each complete.or-partial replacement of the HEPA filter bank.or after anyl structural maintenance on the system housing which could-affect'the-HEPA filter bank bypass < leakage.

i c.

Halogenated hydrocarbon testing shall be performed after each complete or' partial replacement'of-the charcoal adsorber bank or after;any structural maintenance-on the system housing which could effect the'. charcoal adsorber'-

bank bypass leakage.

d.

Each AH-E18A and B (AH-F3A and B)(fan / filter circuit shall' be operating at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month..

i 4.12.1.3 At least once per refueling interval, Lauto'matic initiation.of1 I

the Control Building isolation and recirculation Dampers AH-D28, 37, 39, 'and 36 shall be demonstrated as: operable.

4.12.1.4 An air distribution test shall be. performed on the HEPA filter '

l bank initially, and after any maintenance or testing that could affect :the air distribution within the' system. The: air :

distribution across..the HEPA filter bank shall _be uniform ~

3 within 20%. The test shall be performed at 40,000 cfm -( 10%)l flow rate.

AmendmentNo'.55,ES,fIdd,L175 1

4.12.2 REACTOR BUILDING PURGE AIR TREATMENT SYSTEM Aeolicability: Applies to the reactor building purge air treatment system and associated components (Reference 1).

Obiective:

To verify that this system and associated components will-be able to perform its design functions.

Snecification 4.12.2.1 At least once per refueling interval, it shall be demonstrated that the pressure drop'across the combined HEPA filters and charcoal adsorber banks is less than 6 inches of water at system design flow rate ( 10%).

4.12.2.2 a.

The tests and sample analysis required by Specification-3.15.2.2, shall be performed. initially, once per refueling interval, or within 30 days prior to the movement of irradiated fuel in containment and following significant painting,= steam, fire, or chemical release in any ventilation zone communicating with the system that could

~

contaminate the HEPA filters or. charcoal adsorbers, b.

00P testing shall be performed after each complete or partial replacement of a HEPA filter bank or after-any structural maintenance on the system housing which could affect HEPA' frame bypass leakage, c.

Halogenated hydrocarbon testing shall be performed after each complete or partial replacement of.a' charcoal adsorber bank or after any structural maintenance on the system housing which could affect the charcoal adsorber-bank bypass. leakage.

d.

The DOP and halogenated hydrocarbon testing shall b_e performed at the maximum available flow considering physical restrictions, i.e., purge valve position, and

~

gaseous radioactive release criteria.

e.

Each refueling, AH-E7A&B shall be shown to operste within.

5000 cfm of design flow. (50,000 cfm) with purge valves fully open.

4.12.2.3 An air distribution test shall be performed on the HEPA filter bank initially and after any maintenance or testing that could affect the air distribution within the system. The air-distribution across the HEPA filter bank shall be' uniform within 20%. The. test shall be performed 'at 50,000 cfm ( 10%)

flow rate with purge valves fully open.

t 4-55b Amendment No'. S, Q, jM, Up, IM,175 V

I

~4.17 SHOCK SUPPRESSORS (SNVBBERS):

.y h.

. SURVEILLANCE RE0VIREMENTS 4.17 1 Each snubber shall be demonstrated OPERABLE by performance of the:

following inspection program.

a.

Snubber Tvoes As used in this specification, type of snubber shall mean

^,

snubbers. of the same design and manufacturer, irrespective of p

capacity, b.

Visual Inspections h

Snubbers are categorized as inaccessible or accessible duringL n

reactor operation and may be treated independently. The-TMI l Manager,. Radiological Controls, will ensure that a review is.

performed for ALARA considerations on all snubbers which are o'

located in radiation areas for the determination = of their -

accessibility.

This1 review'shallibe in accordance with the

~

recommendations of Regulatory Guides 8.8 and 8.10.. The-determination shall be based upon the known'or projected -

a radiation levels at each snubber location which would render the area inaccessible during reactor operation and based upon the expected time to perform the visual inspection. - Snubbers ~may L also be determined to~ be inaccessible because of their physical location due to an existing industrial' safety hazard at the specific snubber locatior.. This-determination"shall be reviewed and approved by the Supervisor = of Safety and: Health.

Snubbers ~ accessible during reactor operation shall be' inspected 1.,

in accordance with,the schedule' stated below. Snubbers scheduled for inspection that.are. inaccessible (during reactor operationL because of' physical location or radiation levels'shall:be-inspected during.the next reactor shutdown greater than'48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> where access-is restored

• unless:previously; inspected in s

accordance with the schedule. stated below.'

Visual inspections shall include alle safety 1related : snubbers and shall be. performed in accordance with the followingl schedule:.

.No. Inoperable Snubbers of E'ach Subsequent Visual.

Tvoe Der Inspection Period'

' Inspection Period **#

0 24 months 25%-

1

.16 months:1 25%

2 6-months-25%

l 3, 4 124 days

-1 25%

5,6,7 62 days 25%:

'8 or more 31 days '

25%

Snubbers may continue to be inaccessible'during reactor. shutdown greaters than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> (e.g. if purging of.the reactor building.isinot permitted).

. The inspection' interval; for each type 'of snubber.shall not be lengthened j'

.more than:one, step at a time:unless a generic' problem has beentidentified-

=and corrected; in that event thefinspection interval may be:lengthenedi

'one; step the: first' time and'two steps thereafter~ if. no inoperable 1

Tsnubbers of that type are found.-

q The; provisions. of Table 1.2 are; not ' applicable.

o'

AmendmentNo.79.,JPE,$d,f175.

4 l

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