ML20003B034

Tech Specs 15.3.5 for Instrumentation Sys & 15.6.3 Re Facility Staff Qualifications,Tables 15.3.5-1,5-2,5-3,5-5, 15.4.1-2,1-2,15.6.2-2 & Pages 15.3.1-2,1-3,1-3A for Pressurizer Safety Valves

ML20003B034
Person / Time
Site:	Point Beach
Issue date:	02/04/1981
From:	WISCONSIN ELECTRIC POWER CO.
To:
Shared Package
ML20003B029	List: ML20003B028 ML20003B034
References
TAC-12271, TAC-12272, TAC-43742, TAC-43743, TAC-46520, TAC-46521, NUDOCS 8102100247
Download: ML20003B034 (16)

Text

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, 3. Pressurizer Safety Valves a.

At least ^ne pressurizer safety valve shall be operalile whenever the reactor head is on the vessel.

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b. Both pressurizer safety valves shall be operable whenever the reactor is critical.

4.

Pressurizer Power Operated Relief Valves (PORV) and PORV Block Valves.

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a.

Two PORVs and their associated block va) vet shall be operable.

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1. If a PORV is inoperable, the PORV shall be restored to an . f I

operable condition within one hour or the associated block valve shall be closed.  !

2. If a PORV block valve is inoperable, the block valve shall I

be res.ored to an operable condition within ane hour or the block valve shall be closed with power ramoved from the block valve otherwise the unit shall be shutdown and.in. hot standby within the next six hours. .

5. The pressurizer shall be operable with at least 100 KW of pressurizer heaters available and a water level greater than 10% during steady state power operation. At least one bank of' pressurizer heaters shall' be supplied by an emergency bus power supply.

Basis:

When the boron concentration of the ' reactor coolant system is to be reduced the process must.be uniform to prevent sudden reactivity; changes -in the .  ;

e reactor. Mixing of the reactor coolant will. be sufficient 'to maintain a  ;

uniform boron concentration if at least one' reactor coolant pump or one residual heat removal pump is running while the change is taking place.

.The residual heat removal pump. will circulate the primary system volume'

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8I in approximately one half hour. The pressurize'r is of little concern 20102JQQ f.}

becausa of the low pressurizer volume and bectuce pressurizer boron concentration normally will be higher than that of the rest of the reactor coolant.

Part 1 of the specification requires that a sufficient number of reactor coolant pumps be operating to provide core cooling in the event that a loss of flow occurs. The flow provided in each case will keep DNBR well above 1.30 as discussed in FFDSAR Section 14.1.9. Therefore, cladding damage and release of fission products to the reactor coolant will not occur. Heat transfer analyses (1) show that reactor heat equivalent to 10% of rated ~ power can be removed with natural circulation only; hence, the specified upper limit of 1% .

rated power without operating pumps provides a substantial safety ' actor.

Each of the pressurizer safety valves is designed to relieve 288,000 lbs.

4 per hr. of saturated steam at setpoint. Below 350*F and 350 psig in the reactor coolant system, the residual. heat removal system can remove decay heat and thereby control system temperature and pressure. If no residual heat is removed by any of the means available, the amount of steam which could be-generated at safety valve rellei pressure would be less than half the valves' capacity.

One. valve therefore provides adequate defense against over-pressurization. Part 1 c(2) permits an orderly reduction in power if a' reactor coolant pump is lost during ' operation ~between 10% and 50% of rated power.

Above 50% power, an automatic-reactor trip will occur if either pump is lost.

The power-to-flow ratio will be maintained equal to or less than 1.0 which.

ensures that the minimum DNB ratio increases at lower flow since the maximum o

enthalpy rise does not. increase above .its normal full-flow ' maximum value. (2)

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A PORV is defined as OPERABLE if leakage past > the valve is less than that allowed in Specification 15.3.1.D and the PORV has met its most recent

.R channel test as specified in Table-15.4.1-1. The PORVs operate'to relieve, in a controlled manner, reactor coolant system pressure increases below

^15.3.1-3'

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> .' l the setting of the pressurizer safety valves. These.PORVs have remotely 4

operated block valves to provide a positive shutoff capability should a ,

PORV become inoperable.

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I The . requirement that 100 KW of pressurizer heaters and their associated i controls be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized.during a loss- ,

f of offsite power condition to maintain pressure control and natural cir--

culation at hot standby.

I Re ference (1) PSAR Section 14.1.6

. (2) FSAR Section 7.2.3 t

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15.3.5 INSTRUMEliTATION SYSTEM Operational Safety Instrumentation Applicability:

Applies to plant instrumentation systems.

Objectives:

To provide for automatic initiation of the Engineered Safety Features in the event that principal process variable linits are exceeded, and to delineate the conditions of the plant instrumentation and safety circuits necessary to ensure reactor safety.

Specification:

A. The Engineered Safety Features initiation instrumentation setting limits shall be as stated in Table 15.3.5-1.

B. For on-line testing or in the event of a sub-system instrumentation channel failure, plant operation at rated power shall be permitted to continue in accordance with Tables 15.3.5-2 through 15.3.5-4.

C. In the event the number of ' channels of a particular sub-system in service falls below the limits given in .the column entitled Minimum Operable Channels, or Minimum Degree of Redundancy cannot-be achieved, operation shall be limitr.d according to.the requirement shown in Tables 15.3.5-2 through-15.3.5-4, Operator Action;when minimum operable channels unavailable.

D. The accident monitoring instrumentation channels in Table 15.3.5-5 shall be operable. In the event the number of channels in a parti- i

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cular sub-system falls below the. minimum number of- operable channels  !,

p given in Column 2, operation and subsequent operator action shall - .;_

I be in'accordance with Column 3. l-Basis:

Instrumentation has been provided to ' sense accident . conditions and to-initiate operation of. the Engineered Safety Features (l) .

15.3.5-1

which tutomatically initistas sppropriata action to pravant excasding established limits. Safety is not compromised, however, by continuing opera-tion with certain instrumentation channels out of service since provisions were made for this in the plant design. This specification outlines limiting conditions for operation necesssary to preserve the effectiveness of the l Reactor Control and Protection System when any one or more of the channels is out of service.

Almost all reactor protection channels are supplied with sufficient redundancy to provide the capability for channel calibration and test at power. Exceptions are backup channels such as reactor coolant pump breakers. The removal of one trip channel on process control equipment is accomplished by placing that I

channel bistable in a tripped mode; e.g., a two-out-of-three circuit becomes a or.a-out-of-two circuit. The source and intermediate range nuclea- instrumenta-tion system channels are not intentionally placed in a tripped mode since.these are one-out-of-two trips, therefore the trips sre bypassed during testing.

Testing of the NIS power range channel requires bypassing the Dropped Rod protection frem NIS, for the channel being tested. However, the Rod Position-System still provides the dropped-rod protection. Testing does not trip the system unless a trip condition exists in a concurrent channel.

The operability of the accident monitoring instrumentation ensures that sufficient information is available in selected plant parvaeters to monitor and '

i assess these variables during and following an accident. The PORV block. valves have local, external indication of whether the block valve is open or shut.

If necessary, this local indication can be visually verified during. a .contain-ment entry inspection to verify the block valve is shut.

If the process computer, which provides the reactor coolant system subcooling margin monitor, becomes inoperable, subcooling will be monitored by 'neans of a backup ' plotter medbod or manually using control board instrumentation and

-a saturation curve. _,

15.3.5-5

Reference (1) FSAR - Section 7.5 (2) FSAR - Section 14.3 I (3) FSAR - Section 14.2.5 i

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15.3.5-6

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TABLE 15.3.5-1 . .

ENGINEERED SAFETY FEATURES INITI ATION INSTRUMENT SETTING LIMITS .

NO. FUNCTIONAL UNIT CIIANNEL SETTING LIMIT 1 High Containment Pressure (Hi) Safety Injection * < 6 psig

'2 liigh Containment Pressure (111 -111) a. Containment Spray < 30 psig

b. Steam Line Isolation of Both Lines >

< 20 psig

'3 Pressurizer Low Pressure Safety Injection * > ,1715 psig

_, 4 Low Steam Line Pressure Safety Injection * > 500 psig Lead Time Constant > 12 seconds Lag Time Constant [2 seconds 5 liigh St' e am Flow in a Steam Line . Steam Line Isolation of d/p corresponding to Coincident with' Safety Injection and Affected Line_ <0.66 x 106 lb/hr at Low T,yg 1005 psig

> 540*F 6- High-high Steam Flow in a Steam Line Isolation < d/p corresponding to Steam.Line Coincident with of Affected Line' 4'x 106 lb/hr at

-Safety Injection 806 psig 7  : Low-low Steam Generator Water _ Auxiliary Feedwater > 5% of narrow range Level Initiation instrument

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8' LUndervoltage on 4 KV Busses Auxiliary Feedwater > 75% of normal Initiation voltage

• Initiates also containment isolation, feedwater line isolation _ and starting of all containment fans.

d/p ~ means dif ferential pressure

TABIE 15.3.5-2 (Cont'd) -

1 2 3 4 5' ,

NO. OF MIN. MINIMUM PERMISSIBIE OPERATOR ACTION i

NO.OF CIIANNELS OPERABLE DEGREC OF BYPASS IP CONDITIONS OF NO. FUNCTIONAL UNIT CIIANNELS TO CllANNEIS REDUNDANCY CONDITIONS COLUMN 3 OR 4 t TRIP CANN(yr BE MLT ,

11. Turbine Trip 3 _2 2 1 Maintain <50% of rated power

12. Steam Flow - Feed Water Flow 2/ loop 1/ loop 1/ loop 1/ loop Maintain hot mismatch shutdown

13. Io'Io Steam Generator- 1/ loop 2/ loop 2/ loop 1/ loop Maintain hot

' Water Level- shutdown

14. Undervoltage-4 KV Bus 2/ bus 1/ bus 1/ bus' --

Mainain hot (both buses) shutdown 15.' Underfrequency 4 KV Bits 2/ bus 1/ bus 1/ bus --

Maintain hot (both buses) shutdown NOTE'l': ' When block condition exists, mintain normal operation.

F.P. a Full Power-

• 'Not Applicable

• • On'e additional cl.annel 'may be taken out of service for zero power physics testing.

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TABLE 15.3.5-3

• EMERGENCY COOLING .

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1 2 3 4 5 OPERATOR ACTION <

NO. OF MIN. MIN. PERMISSIBLE IF CONDITIONS OF NO. OF CIIANNEIS OPERABLE DEGREE OF BYPASS COLUMN 3 OR 4 NO. FUNCTIONAL UNIT C11ANNELS M TRIP CIIANNEIS REDUNDANCE CONDITIONS CANNO'r BE MET 3.- SAFETY INJECTION

o. Manual 2 1 1 1 Ilot Shutdown ***

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liigh Containment Pressure 3 2 2 1 Ilot Shutdown'* ,

@. Steam Generator Low Steam Primary Pressure is Pressure / Loop 3 2 2 1 Less than 1800 psig flot Shutdown * **

G._ Pressurizer Iow Pressure 3 2 2 1 Primary Pressure is Less than 1800 psig Ilo t Shutdown * * *

2. CONTAINMENT SPRAY

s. Manual 2 2 2 - ** Ilot Shutdown ***

@. Hi-Hi Containment Pressure 2 sets 2 of 3 2 per 1/ set flot Shutdown * * *

(Containment Spray) of 3 in each set set i3.' AUXILIARY FEEDWATER

.a. , Low-low Steam Gencr' tor Water

i. Start. Motor Driven Pump' 2/ steam gen. 2/either gen. 2 1 Hot Shutdown ***

.il. Start Turbine' Driven ~3/ steam gen. 2/both gens. 2 1 flot Shutdown * * *

' Pump

b.- Trip of both Main Feedpumps' starts motor driven pumps 2/ pump 1/ pump- 1/ pump 1 Ilot Shutdown * *

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c.- Undervoltage on 4KV Busses' starts Turbine driven pump 2/ bus 1/ bus 1/ bus -- flot Shutdown ***

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TABLE 15.3.5-3 EMERGENCY COOLING 1 2 3 4 5 OPERATOR ACTION NO. OF MIN. MIN. PERMISSIBLE IF CONDITIONS OF NO. OF CIIANNEIS OPERABLE- DEGREE OF BYPASS COLUMN 3 OR 4 FUNCTIONAL UNIT CilANNELS TO TRIP CliANNELS REDUNDANCY CONDITIONS CANNOP BE MLT AUXILIARY FEEDWATER(Continued)

Safety . Injection Signal .S.I. INITIATING FUNCTIONS.AND REQUIREMENT AS IN 1. ABOVE Starts Motor Driven Pumps page.2 of 2 P

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TABLE 15.3.5-5 .,

INSTRUMENT OPERATING CONDITIONS FOR INDICATION '

1 2  :

FUNIMUM 3 i NO. OF OPERABLE OPERATOR ACTION IF CONDITIONS NO. FUNCTIONAL UNIT CllANNELS CilANNEL OF COLUMN 2 CANNOT BE MET

1. PORV Position Indicator 1/ Valve 1/ Valve If the operability of the PORV position J adicator cannot be restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, shut the associated PORV Block Valve.

2. PORV Block Valve Position Indicator 1/ Valve 1/ Valve If the operability of the PORV Block Valve Position Indicator cannot be restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, shut and verify the Block Valve shut by direct observa-tion or declare the Block Valve inoperable. l

3. Safety Valve' Position Indicator 1/ Valve 1/ Valve If the operability of the Safety Valve Position Indicator cannot be restored within seven days, be in at least flot Shutdown witbin the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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4. . Reactor Coolant System Subcooling 1 1 If the operability of a subcooling monitor cannot  !

be restored or a backup monitor made functional  !

within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, be in at least flot Shutdown within '

the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

5. Auxiliary Feedwater Flow- Rate

• 1 0 If auxiliary feedwater flow indication becomes ,

inoperable while auxiliary feedwater is required  ;

for the Steam Generators, the Steam Generators i should be monitored to establish adequacy of auxiliary feedwater flow.

6. Control Rod Misalignment as Monitored 1 1 Log individual rod positions once/hr. , af ter a load

.by On-Line Computer change >10% or after >30 inches of control rod motion.

• Applies to presently installed auxiliary feedwater pump discharge flow indicators.  ;

TABLE 15.4.1-1 (CONTINUED)

Chrnnel Description Check Calibrate Test Remarks 24.- Containment Pressure S R M** Narrow range containment pressure

(- 3. 0, +3 psig excluded)

25. Steam Generator Pressure S*** R M***

26. Turbine First Stage Pressure S** R M**

27. ' Emergency Plan Radiation M R M Instruments 28.. Environmental 14onitors M N.A. N.A.

29. .Ovsrpressure Mitigating S R ****

30. PORV Position Indicator N.A. R N.A.

31. PORV Block Valve Position Q R N.A.

Indicator

32. Szfety Valve Position Indicator M R N.A.

33. PORV Operability N.A. R M Performance of a channel functional test but excluding valve operation.

34. Subcooling Margin Monitor M R N.A.

35. Undervoltage on'4KV Bus. N.A. -R M** For Auxiliary Feedwater Pump Initiation

36. Auxiliary Feedwater Flow Rate See Remarks R N.A. Flow Rate indication will be checked at each unit startup and shutdown S - Each Shift M - Monthly D - Daily P - Prior to each startup if not donc previous week.

W - Weekly -R - Each Refueling Shutdown (Bist not to exceed 20 months) .

Q - Quarterly' N.A. - Not applicable.

2/W - Biweekly

i TABLE 15.4.1-1 (CONTINUED)

• • Not required during periods of refueling shutdown, but' must be performed prior to starting up if it has not. been

. performed during the previous surveillance period.

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r ooo Not required 'during periods of refueling shutdown if steam generator vessel temperature is greater than 70*F.

• • • • 'When used 'or the overpressure mitigating system each PORV shall be demonstrated operable by: I a ., Performance of a channel functional test on the PORV actuation channel, but excluding valve operation,

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i within 31 days prior -to entering a condition in which the PORV is required operable and at least once

- perf 31 days . thereaf ter when the PORV is : required operable.

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is. . Testing valve operation in accordance with the inservice -test requirements of the ASME Boiler and Pressure Vessel Code,Section IX.

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TABLE 15.4.1-2 (CONTZi6UED)

Test Frequency

14. Refueling System Interlocks Functioning Each refueling shutdown

15. Service Water System Functioning Each refueling shutdown

16. Primary System Leakage Evaluate Monthly (6)

17. Diesel Fuel Supply Fuel inventory Daily

18. Turbine Stop and Functioning Monthly (6)

Governor Valves

19. Low Pressure Turbine Visual and magnetic Every five years Rotor Inspection (5) particle or liquid penetrant

20. Boric Acid System Storage Tank Daily Temperature

21. Boric Acid System Visual observation Daily of piping temperatures

-(all 3,145'F)

22. Boric Acid Piping Heat Electrical circuit Monthly Tracing operability-

23. PORV Block Valves Complete Valve Cycle Quarterly (6)

24. Integrity of Post Accident Evaluate. Yearly Recovery Systems Outside Containment (1) A radiochemical analysis for this purpose shall consist of a quantative measure--

ment of each radionuclide with half life of >30 minutes such that at least 95%

of total activity of primary coolant is accounted for.

(2) E determination will be started when the gross activity _ analysis of a filtered sample indicates 1,10 pc/cc and will be redetermined if the primary coolant gross radioactivity of a filtered sample increases by more than 10 uc/cc.

(3) Drop tests shall be conducted _at rated reactor coolant flow. Rods shall be dropped under both cold and hot conditions, but cold drop tests need not be timed.

(4) Drop tests will be conducted in the hot condition for' rods on which maintenance-was performed. ~

(5) As accessible without disassembly of rotor. .

(6) Not required during periods of refueling shutdown.

(7) At least once per week during periods of refueling shutdown. T (8) At least three times per week (with maximum- time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> between samples)

. during periods of refueling shutdown.

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15.6.3 Frcility Staff Qualifications e

15.6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions or as clarified in 15.6. 3. 2 thrcagh 15.6. 3.4.

15.6.3.2 Except as provided in 15.6.3.3, either the Radiochemical Engineer or the Health Physicist shall meet the following requirements:

a. The individual shall have a bachelor's degree or the equivalent in a science or engineering subject, including some formal training in radiation protection. For purposes of this paragraph, " equivalent" is as follows:

(1) Four years of formal schooling in science or engineering; or (2) Four years of applied radiation protection experience at a nuclear facility; or (3) Four years of operational or technical experience or training in nuclear power; or (4) _ny combination of the above totallino four years.

b. Except as provided in d. , below, the individual shall have at least five years of professional experience in applied radiation protection. A master's degree in a related' field is equivalent to one year of experience and a doctor's degree in a related field is equivalent to two years of experience.

c. Except as provided in d., below, at least three of the five years of experience shall be in applied radiation protection work in a nuclear facility dealing with radiological problems similar to those encountered in nuclear power plants.

d. If the individual has a bachelor's degree specifically in health physics, radiological health, or radiation protection, at least three years of professional experience is required; if the' individual has a master's or a doctor's degree specifically .in health physics, radiological health, or radiation protection, at least two years of professional experience is required. This experience shall be in applied radiation protection in a nuclear facility dealing with radiological problems similar to those encountered in nuclear-power plants.

15.6.3.3 In the event the position.of Radiochemical Engineer or Health Physicist is vacated and neither the remaining individual nor the proposed replace-ment meets the qualifications of 15.6.3.2, but one of these indiriduals is determined to be otherwise well qualified, then concurrence of NRC shall be sought in approving the qualification of that individual.

15.6.3.4 The Duty Technical Advisor shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in' plant 8 design and response and analysis of the plant for transients and accidents.

15.6.3-1

Manz.ger - (1)

CONDUCT OF PLANT OPERATIONS tiuclear Operations

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FIGURE 15.6.2-2 ,

Manager's Supervisory Staff Committee Assistant to (1) l the Manager l l

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OPERATIONS TRAINING MAINTEtiANCE & TECHNICAL SERVICES l CONSTFUCTICN l l

Superintendent - (1) Training (1) Superintendent (1) Superintendent (1) l Operations (FRO) Supervisor Maintenance & Technical Services  !

Qi M STRY &

Construction INSTFUMEha g e

& CONTROL REACTOR ENGINEERING .

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Technical (1) Instrument & Reactor Radiochemical 4 Aesistant Engineer (1) l Control Eng. (1) ] Engineer (1) l l l

. Shift Supervisor (SRO) Maintenance Health (1) .j On2 per shift Supervisors Technical Physicist Technicians Assistants Operating Supervisor Duty Technt- Technical One per shift (SRO or RO cal Advisor Assistants Notes:

1 Per Shift Note 4 . 1. The Operations Group shift makeup is the minimum size for operation in all Ccntrol Operator (RO) modes except cold shutdown of Point Beach Unit Nos. I and 2. The Oper-On2 per shift-one unit ations Group shift makeup may be less than the minimum requirements for a Two per shift-two units period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members, provided immediate action is taken to restore the shift makeup to within the minimum requirements.

Auxiliary Operator 2 In administrative matters and routine HP concerns, the !!P reports directly to Two per shift-one unit Three per shift-two units the Radiochemical Engineer .n matters of radiological health and safety policy determination, interpretation, or implementation, the HP (based on IIP iudge-i ment) reports directly to the Manager-Nuclear Operations.

3. SRO - NRC Senior Reactor Operator License RO - NRC Reactor Operator License

4. The Duty Technical Advisor is located on site on 10 minute call to the control room. Unexpected absences will be treated as in Note 1.

4.,