ML19305B343

Supplements 791231 Ltr Re Lessons Learned Task Force Implementation.Includes Info Re Emergency Power Supply, Direct Position Indication of Relief & Safety Valves & Instrumentation for Inadequate Core Cooling

ML19305B343
Person / Time
Site:	Prairie Island
Issue date:	03/13/1980
From:	Mayer L NORTHERN STATES POWER CO.
To:	Office of Nuclear Reactor Regulation
References
RTR-NUREG-0578, RTR-NUREG-578 TAC-12428, TAC-12429, NUDOCS 8003190554
Download: ML19305B343 (56)

Text

$f I q NSF NORTHERN STATES POWER COMPANY MIN N E A PO L.l a. M IN N E S OTA 55409 March 13, 19 80 Director of Nuclear Reactor Regulation U S Nuclear Regulatory Commission Washington, DC 20555 -

PRAIRIE ISLAND NUCLEAR GENERATING PLAfff Docket No. 50-282 License No. DPR-42 50-306 DPR-60 1/1/80 Lessons Learned Implementation Additional Information Attachment I supplements our December 31, 1979 letter on Lessons Learned implementation.

L 0 Ma e ager of Nuclear Support Services LOM/J AG,'ak cc: J G Keppler G Charnoff I l

At tachme nt l l

8003190554

t' e ATTACHMENT 1 to NSP LETTER DATED MARCH 13, 1980 TABLE OF CONTENTS Item Page 2.1.1 Emergency Power Supply 1 2.3.3.a Direct Position Indication of Relief & Safety Valves 7 2.1.3.b Instrumentation for Inadequate Core Cooling 15 2.1.4 Diverse Containment Isolation 27 2.1.5.a Dedicated H 2 Control Penetrations 28 2.1.6.a System Integrity for High Radioactivity 30 2.1.5.b Plant Shielding Review 32 2.1.7.b Auxiliary Feedwater Flow Indication to Steam Generator 36 2.1.8.a Post Accident Primary Coolant System and Containment 37 Atmosphere Sampling System 2.1.8.b Increased Range of Radiation Monitors 40 2.1.8.c Improved In plant Radio Iodine Monitoring 43 l 2.1.9 Reactor Vessel Head and Pressurizer Venting 44 2.2.1.a Shif t Supervisor Responsibility 45 2.2.1.b Shift Technical Support 46 2.2.1.c Shift & Relief Turnover Procedure 48 2.2.2.b Onsite Technical Support Center 49 2.2.2.c Operational Support Center 54

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I 2.1.1 EMERGENCY POWER SUPPLY Our response dated December 31, 1979, stated: ". . . by a manual transfer switch located in the Rod Drive Power Supply Room which is located within 1 minute of the Control Room." This has been confirmed by timing from op erator in control room being directed to perform transfer, walking to the transfer location and performing transfer. Total clapsed time was 1 minute 17 seconds.

We are not aware of any analysis which specifically defines the time frame for needing pressurizer heaters. Westinghouse has stated: "The energy of the steam and water contained in the pressurizer assures that pressure will be maintained to prevent bulk boiling in the core for a sufficient period of time following reactor trip to permit restoration of power to the heaters."

Our position is that the heaters are useful in maintaining natural circula-tion during a loss of offsite power event or as a subsequent -action stabili-zation mechanism during smaller high energy line breaks. For these purposes they are not necessary until sometime beyond 10 minutes from event initiation.

They would be useful as long as natural circulation is used to cool the reactor-to 350*.

Safeguards bus 120 and 220, Unit 1 and Unit 2 respectively, contain the safe-guards ' source breaker for backup supply to group B heaters. The buses and breakers are safety-grade. (See Figure 86 3-1)

By test at refueling shutdown conditions (cold and depressurized), the PORV's were cycled 15 times. This will significantly increase when hot and pressurized since the driving force of system pressure will assist the air opeator in over-coming the closing spring.

Attached Figures 7.2-4, 7.2-5, 7.2-11 (Sheet 1) and 8.3-3 from the Prairie _ Island FSAR illustrate the power source and instrument channel redundancy for 'Jnit 1.

Unit 2 is identical in design; bus and panel identification numbers are different.

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l 2.1. 3. a - DIRECT POSITION INDICATION OF RELIEF & SAFETY VALVES The control room alarm is labeled " Pressurizer Safety / Relief Valve Flow".

Flow rates and acoustic signals do not have a direct correlation since the signal measures the acoustic energy caused by the flow through a valve.

This energy may or maynot be directly proportional to the flow. This could only.be determined by running actual flow test at various pressure differences and measuring the acoustic signal. However, from EPR'I test data, the full flow vibration energy on actual blowdowns is in the 1000 to 2000 g range.

The alarm on the system is set at 40 g's. Therefore the minimum flow rate or leakage rate which can be detected is at between two to four percent of the

-full blowdown flow energy. Because there is no correlation between the flow and the acoustic signal, it can not be determined if the setpoint of 40 g's is above or below the capacity of the plant's charging pump. Using the 1000 to 2000 g value for the acoustic energy for the blowdown on one sensor, the other two sensors would read between 8 to 10 g's which is below the 40 g's alarm setpoint for the system. This cross-talk value was determined from impact level tests on the individual sensors.

The plant operators were informed about the acoustic valve monitoring system by the attached training department update.

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?' L CORRESPONDENCE

. I#525

. onTc February 8,1980

,,e Ban StephenS LOCATION Prairie Island Distribution tocATioN vo;ccT VALVE POSITION MONITORING SYSTEM As a result of' TMI Lessons Learned. A Valve Monitoring System (VHS) has been installed. This system detects and indicates noise produced when a valve is open.

The valves monitored are the Pressurizer Safetics (one sensor for two valves) 'and the' Pressurizer Power Operated Relief Valve (PORV). The system consists of an accelerometer, preamplifier, signal conditioner, and control r on indicating lights and alarms. (Figure 2) .

The (VMS) is physically divided into two locations. (1) All accelerometer sensors and their respective pre-amplifiers are located inside the containment building, and (2) all signal conditioning and alrrm relays are located in the relay room. Outputs from the aignal condition cabinentii25 are routed to thu main control room, i.e.,

3 lamp light module and common annunciator 01-07 on panels C-1, C-2 and 012 or 0512.

Also a digital and analog output is provided to the P-250 computer. Y9201D and Y4111A.

J FUNCTIONAL DESCRIPTION Figure 1 shows the front panel controls and indicators on the signal conditioners located in the signal conditioning cabinet 125 in the relay room. Unit #1 is the top module and Unit #2 1s the bottom module. Each module holds three channels of signal conditioner's. Reading left to right are Chl, Ch2, Ch3, 2 blank panels and clarm reset push. buttons in the last panel. From top to bottom the controls and indicator for each signal conditioner are as follows:

1. Meter - Displays the ras acceleration value of the input signal in g's (assumed to be sinusoldal) -in terms of percent full scale as selected by the FULL SCALE RANGE switch. .

2. Power indiento,r _ A green lamp located d!.rectly below the meter. It is 11-luminated when power is applied. NOTE: The VMS is a fail-safe system therefore, jp . a loss of power' vill activate all lamp nodule lights & control room annunciator.

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3. . HI-ALARM indicator - A red lamp at the left of the power indicator. It is 11- r luminated when a'high-alarm condition has occurred and remains on until the external push button (reset) in last panel of Unit #1 or 2 nodule (for as- .

sociated channel) is depressed. This push button also will reset (the as-sociated channel) lamp light in lamp module and allow's the reset of the com-mon alarm annunciator 01-07 in the main control room (i.e. , C-1, C-2 and 012 or 1

0512).

Upon receipt of a.Hi alarm, verify which channel or channels caused the alarm.

If more than one channel is off-scale high, up the FULL SCALE RANGE switch (15) until a meter indication is received. Note the signal level as read on (1) meter. Compare other redundant sources of valve open indication (i.e., temp, press light indication). Confirmation is.that the valve is open then refer to establish procedures for the open valve.

4. LO-ALARM indicator - A red lamp at the right of the power indicator. It is 11-luminated when a low-alarm condition has occurred and remains on until the ex-tornal push button (reset) in last penel of module of Unit #1 or 2 (for associated channel) is depressed. For continuing low-level signals accompanied by LO-ALARM i

(4) Icd light "0:i" at signal conditioner unit. Determine fault by depressing the read PRE-AM BIAS (11) and observed the output meter (1) > 80% indicates faulty (open cable) between signal conditioner and pre-amplifier, or faulty pre-amplifier. ' 20% indicates rhorted or partially shorted cabic between signal conditioner and pre-arplifier. Correct adjustr.ent of the bias will in-dicate = 55%. Note that this alarm will not activate any control room devices

! and the "dereat switch (6) will be norrally be in the defeat position (sec #6) t l k ). DoloW-l.

9

I#525 2/8/80 Page 3

5. HI-ALARM DEFEAT .- When this switch is positioned upward, the alarm system is operational and when it is positioned down, in the DEFEAT position, the alarm circuit functions normally; however, the alarm output signal is de-energizes the

- output relay. This causes the alarm functions to operate, lighting the control room light module and alarm.

6. LO-ALARM DEFEAT - When this switch is positioned upward, the alarm system is operational and when it is positioned down, in the DEFEAT' position, the alarm circuit functions normally, however, the alarm output signal is not trans-mitted to the alarm output connector. No output relay provided.

7. SENSITIVITY DIAL - Ten-turn locking dial calibrated in terms of transducer sensitivity in picocoulombs per g (pC/g). The number in the window is the first significant figure, the number engraved in the outer dial is the second significant figure, and the third significant figure can be extrapolated between N the five divisions of the numerals engraved in the outer dial. The outer dial k

numbers should be read relative to the mark located immediately below the window displaying the first significant figure. A locking knob is located in the upper right-hand corner of the SEUSITIVITY dial. Moving the knob up releases the locking mechanism and moving it down locks the dial to preclude inadvertent changes. A range of 5 to 15 pC/g is provided.

8. HP, IN/0UT - Two-position toggle switch used to activate (IN) or bypass (OUT) the three-pole active high-pass filter. This is to be in the IN position

9. LP, IN/0UT - Two-position toggle switch used to activate (IN) or bypass (0UT) l the three-pole active low-pass filter. This is to b2 in the IN position

10. READ HI-ALARM level]2 - With this button depressed the high-alarm level is dis-played on the panel meter.

11. READ PREAMP BIAS - While this button is depressed the remote charge preamplifier R

3 bias is displayed on the panel meter. The voltage is displayed as a precent of full scale with 30 V corresponding to 100%. An acceptable bias level is 9 to 24 V dc - this rar:ge corresponou to a cetur reading of 30 to 807. of full scale.

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L#525 2/8/80 l f'N Page 4

12. TEST HI-ALARM - Depressing this button triggers the high-alarm circuits for test purposes. This test circuit also will illuminate the associated' channel light on three lamp assembly in Cl or C-2 and cause annunciator in 01-07, 012 or 0512 to illuminate. The digital input to the P-250 computer will indicate a change of state. To clear the test circuit depress " reset" push button in last panel in associated module for channel under " test".

13. HI-ALARM level - This six-position, screwdriver-operated switch selects the alarm level in steps of 20, 30, 40, 60, 80 and 100% of the FULL SCALE RANGE, level.

14. PREAMP BIAS ADJUST - This screwdriver-operated control adjusts the preamplifier bias voltage.

15. FULL SCATJ. RANGE - This screwdriver-operated switch adjusts the full-scale out-A g put range in six steps. The FULL SCALE RANGE is adjustable from 1.0 to 300 g rms.

CALIBRATION AND MAINTENANCE The signal conditioners and pre-amps are solid-state devices and do not require frequent calibration. The high and low filter break frequences are set at time of initial system calibration and should not require periodic adjustments. NOTE: Unit

1. 1 has received the. initial calibration, Unit #2 will be calibrated prior to February 1, 1980.

POWER SUPPLY Unit #1 is supplied 120 VAC power from Distribution panel 1117 breaker #10. To plug-mold identified by #1. The 3-point power cabic for Module #1 should be plug into the above power source.

11

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,,s. I#525

,5 2/8/80

Page 5 Unit #2 is supplied 120 VAC power from distr'ibution panel 217 breaker #6. To plug-mold identified #2. The 3-point power cable should be plugg dd'into.the above power source for Module #2.

Unit #1 Lamp Box Assy Unit #2 Lamp Box Assy Top ch 1 Safties Top ch 1 Safties Middle ch 2 CV-31231 Middle ch 2 CV-31233 Botton ch 3 CV-31232 Bottom ch 3 CV-31234 Sincerely, g

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/s/ B F Stephens B. F. Stephens Assoc. Prod. Engineer BFS/jd cc: Chron File Subj File - Reactor Coolant Sys. (Gen.)

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2.1.3.b INSTRQ3ENTATION FOR INADEQUATE CORE COOLING A safety-grade subcooling margin monitor is installed at Prairic Island, on both units. The temperature input is multiple core exit thermocouples and the pressure input is Reactor Coolant System (RCS) wide range pressure. Wide range pressure must be used, since RCS pressure is likely to fall below 1700 psig (narrow range downscale pressure reading) for most postulated accidents. Core exit thermocouples are used because they indicate actual core conditions rather than loop conditions.

At present, neither wide range pressure nor incore thermocouples can be classi-fled as safety-grade. Both inputs are being upgraded. Safety-grade pressure transmitters are on order with scheduled delivery by July,1980. For Unit 1, separate pressure taps and the safety-grade transmitters will be installed dur-ing the refueling outage currently scheduled for July, 1980. For Unit 2, the pressure taps have already been installed and the transmitters can be installed without an outage, shortly after delivery.

The cabling and racks outside containment must also be upgraded to safety-grade, and it is anticipated that this can be accomplished by January 1, 1981.

Until the inputs to the subcooling monitors are upgraded to safety-grade, the P-250 plant computer has been programmed to accept the best available tempera-ture and pressure inputs as follows:

1. If loop pressure is 1700-2500 psig, narrow range pressure (which is safety-grade) is used.

2. If pressure is less than 1700 psig, wide range pressure is used.

3. If reactor coolant pumps are running, safety-grade RTD's from the hot leg bypass loops is used.

4. If reactor coolant pumps are not running, compensated core exit thermocouples (at least 2 per quadrant) are used, i

Subcooling margin, as determined by the P-250 computer, is displayed on a CRT t in the control room. I l

1 Procedure El.1, Appendix I, " Termination of Safety Injection Following LOCA" has 1 l

been modified by a temporary memo which instructs the operator to use the P-250 for subcooling margin indication until the subcooling margin monitor inputs are upgraded to safety-grade. 15 1

1

, t Attachment to 2.1.3.b NORMAL AMBIENT ACCURACY COMBUSTION ENGINEERING MODEL 001 SUBC00 LED MARGIN MONI'IOR i

j ASSUME: RCS PRESSURE = 2000 psig = 2015 psia 1

-PRESSURE INPUT = PT-419 RCS wide range (0-3000 psig)

PT-420 RCS wide range (0-3000 psig)

TEMPERATURE INPUT = T03 (C-6) INCORE THERMOCOUPLE (QUADRANT 1)

. T36 (.J-9) INCORE TilERM0 COUPLE (QUADRANT 3)

T7 (G-12) INCORE THERMOCOUPLE (QUADRANT 2)

T34 (J-3) INCORE TilERM0 COUPLE (QUADRANT 4) 1 MINIMUM ABSOLUTE PRESSURE =-RCS PRESSURE - MAX. NORMAL PRESSURE ERROR

=.2015 psia - 32 psi = 1982 psia T-SAT FOR 1982 PSIA = 634.7'F (KEENAN AND KEYES STM. TABLES - 1969)

INPUT TEMPATURE_VALUE REQUIRED TO ENSURE SUBC00 LING = T-SAT (MIN. ABS. PRESS.) - MAX NORMAL TEMP.

AT MIN. ABS. PRESSURE ERROR

= 634. 7'F. - 8.19'F (T/C)

= 626.5'F ERROR IN CALCULATED SUBC00 LING MARGIN = T-SAT (RCS PRESS) - T(INPUT VALUE)

= 637.0*F - 626.5'F

= 10.5'F

-ERROR DUE TO DISPLAY DEVICE ERROR = + 1/2% of OUTPUT SPAN (Comb.~Eng. Technical: manual, pg 2-2

.subcooled margin monitor model 001) = (.005) (150*F)

= + .75'F TOTAL INDICATED SUBC00 LING ERROR ' = 10.5'F + .75'F

= 11.25 *F

.16

ny- . ,

Attachment to 2.1.3.b NORMAL AMBIENT ACCURACY INCORE THERMOCOUPLES Normal Thermodouple Accuracy (Ref: Incore Inst. Manual for Prairie Island i 3/8% @ 530'F - 700*F) = 1 3'F

- Reference Junction Accuracy (Ref: Instruction Manual - Model BRJ Reference Junction i .l*F + .2'F) = 1 3*F Calibration Accuracy - Ref Junction (Ref: PI' Calibration Procedure) .= + 3*F

- Maximum Instrument Drif t = 1% of span (From PBNP Calculations) = .001 of 700*F

= 1 7*F MAXIMUM NORMAL THERMOCOUPLE ERROR (without display error) 32 , ,32,32+7

=.1 8.19 *F With Display Errors Included

1. ' Max Error with P250 (11% A to D error)

• 1 10.7 *F

2. Max Error with Subcooling Meter (1/2% of output span, 100-999'F)~ = 1 12.69*F

~ 3. Max Error with Multipoint Indicator

(.2% output span, 400-700*F) = 1 8.79"F Above errors are considered random and independent, and assumed to occur all in the same direction.

17 i

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t- 9 Attachmsnt to 2.'l.3.b 5 .

NORMAL AMBIENT ACCURACY LOOP 419, 420 RCS WIDE RANGE PRESSURE I. Normal Transmitter Reference = .25% of span Accuracy-(includes hysteresis, 1 0025 of 3000 psi Linearity & Repeatability) 1 7.5 psi

~

II Accuracy of Dead Weight Tester-

& Heise gauge to calibrate transmitter

.(DWT used to cal Heise which is used to cal loop)

DWT 1 1% of span 3 psi Heise 1 1% of span i 3 psi =

1 4.2 psi Total / 32+32 =_i 4.2 psi III -Allowed Calibration Tol. = f .5% of span i .005 of 3000 psi i 15 psi

' IV Ambient Temp effect on Transmitter = 1 5% of span (used PBNP error) j; .005 of 3000 psi j; 15 psi V Maximun Transmitter Drif t during Cal = + .25% of span interval. (Review of "as found" data ~+ .005 of 3000 psi shows performance within j .25% of Cal.)

1 7.5 psi CURRENT REPEATER (419 LOOP)

I Normal Inst Accuracy =

1 5% of span

.005 of 3000 psi i 15 psi II . Allowed Cal Tol. = 1 5% of span

.005 of 3000 psi i 15 psi

+

4 18

,- --a- . -- - c - , - - - , , -

.e- .

Attachment to 2.1.3.b Maximum normal instrument error

/ 7.5 + 4.2 + 15 + 152 ,.7,3T} 152 + 15 2

=- + 32 psi ~for' Loop 419

=- 7.5 + 4.22 + 15 + 152'+.7.52

= + 24 psi for Loop l420 Above errors are considered random and independent and are combined by taking the square root of - the sum of the squares.

)

In summary the above analyses includes the errors assumed for

-1) 'Specified instrument' accuracy.

2) Calibration instruments t -3) Allowed Calibration Tol l' '

_4) Ambient Temperature effect

, 5) Transmitter drif t between calibration intervals .

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NORMAL AMBIENT ACCURACY WESTINGHOUSE MODEL P250 PLANT PROCESS COMPUTER SUBCOOLING MONITOR PROGRAM FUNCTION.

4- ASSUME : RCS PRESSURE = 2000 PSIG = 2015 PSIA PRESSURE INPUTS = PT-429 PRESSURIZER NARROW RANGE P(1700 - 2500 PSIG)

PT-430 PRESSURIZER NARROW RANGE PT-431 PRESSURIZER NARROW RANGE

~

PT-419 RCS WIDE RANCE PRESSURE (0-3000 PSIG)

PT-420 RCS WIDE RANGE PRESSURE TEMPERATURE INPUTS = TE-401A RCS BYPASS LOOP RTD (520 - 620 F)

TE-401B RCS BYPASS LOOP RTD TE-402A RCS BYPASS LOOP RTD

-TE-402B RCS BYPASS LOOP RTD TE-403A RCS BYPASS LOOP RTD TE-403B RCS. BYPASS LOOP RTD TE-404A RCS BYPASS LOOP RTD TE-404B RCS BYPASS LOOP RTD TE-405A RCS BYPASS LOOP RTD TE-405B RCS BYPASS LOOP.RTD TE-406A RCS BYPASS LOOP TRD TE_406B RCS BYPASS LOOP RTD

. TE-407A RCS BYPASS LOOP RTD TE-407B RCS BYPASS LOOP RTD TE-408A RCS BYPASS LOOP RTD TE-408B RCS BYPASS LOOP RTD' T01 - T39 RCS CORE EXIT THERMOCOUPLES (0-1200 F)

-MINIMUM AhSOLUTE PRESSURE = RCS PRESSURE = MAXIMUM NORMAL PRESSURE ERROR

= 2015 PSIA - 12 PSI = 2003 PSIA-T-SAT FOR 2003 PSIA = 636.21 F -(KEENAN AND KEYES STEAM TABLES - 1969)

INPUT TEMPERA *;URE VALUE )

REQUIRED TO ENSURE SUBC00 LING ) = T-SAT { MIN. ABS. PRESS.) - MAX NORMAL TEMP ERROR AT MINIMUM ABS PRESSURE ) = 636.21 F - 2 F

= 634.2 F i

b 20

MODEL P250 (con't)

ERROR .IN CALCULATED SUBC00 LING MARGIN = T-SAT (RCS PdESS) - T(INPUT VALUE)

= 637 F - 634.2 F

= 2.8 F TOTAL INDICATED SUBC00 LING ERROR =2.8F-3F POST ACCIDENT ENVIRONMENTAL ACCURACY WESTINGHOUSE P250 PROCESS COMPUTER SUBC00 LING MONITOR PROGRAM FUNCTION ASSUME: RCS PRESSURE = 2000 PSIG = 2015 PSIA USING NARROW RANGE PRESSURE + CORE EXIT THERMOCOUPLES: AS INPUTS CONT. PRESSURE = 60 PSIG STEAM AND TEMPERATURE = 285 F MINIMUM ABSOLUTE PRESSURE = RCS PRESSURE - MAX. POST LOCA PRESSURE ERROR

= 2015 PSIA - 113 PSIA = 1902 PSIA T-SAT FOR 1902 PSIA = 628.9 F (KEENAN AND KEYES STM TABLES - 1969)

- INPUT TEMPERATURE VALUE REQUIRED TO ENSURE SUBC00 LING = T-SAT (MIN. ABS. PRESS'1 - MAX. POST LOCA TEMP ERROR AT MIN. ABS. PRESSURE = 628.9 F - 8.20 F (T/C ERROR)

= 620.7 F ERROR IN CALCULATED SUBC00 LING MARGIN = T-SAT (RCS PRESS) - T INPUT VALUE

= 637 F - 620.7 F

= 16.3 F TOTAL INDICATED SUBC00 LING ERROR = 16.3 F 1

l

a o PRESSURIZER PRESSURE NARROW RANGE NORMAL AMBIENT ACCURACY

'l [2] LT 429, 430, 431 PT (I/I I/I CI (426 & 428)

I Normal Transmitter Reference Accuracy = + .25% of Span (includes Hysteresis, Linearity_and = T .0025 of 800 psi Repeatability) = + 2.0 psi II Accuracy of Dead Weight Tester and Heise Gage to Calibrate Transmitter DW" + .1% of Span (3000) = + 3 psi Heise + .1% of Span (3000) =}3 psi TOTAL 3 +3 = + 4.2 psi = + 4.2 psi III Allowed Calibration Tolerance = + .5% of Span ,

= I .005 of 800. psi

~

= 4 psi IV Ambient Temp Ef fect on Transmitter = + .5% of Span (Used Point Beach Nuclear Plant error) = T'.005 of 800 psi

=

[4 psi V Maximum Transmitter Drift during = + .25% of Span Calibration Interval (Review of as-found = + .0025 of 800 psi data shows performance within + .23% of

~

= I 2 psi

~

calibration) i; ^ 22

o 'o' CURRENT REPEATER let I Normal' Instrument Accuracy = + .5% of Span

= T .005 of 800 pai

=

}[4pai II Allowed Calibration Tolerance = + .5% of Span

= 7 .005 of 800 psi

=

}[4 psi CURRENT REPEATER 2nd

= + 4 psi T. Normal Instrument Accuracy (as above)

II Allowed . Calibration Tolerance = + 4 psi (as above)

COMPUTER INPUT Accuracy .125% of Span =

j;1 psi Repeatable .125% of Span =

j;1 psi Resolution .0625 of Span =

j;.5 psi Maximum Normal Instrument Error

= )22 + 4.22+42.42+22+42 +42+4212g2+52

= j; 11.13 psi In summary the above analysis includes the errors assumed for:

1)- Specified Inst Accuracy 5) Transmitter Drift between Cal Interval ~

2) Cal Inst 6) Current Repeater Ef fects

3) Allowed Cal Tol 7) Computer Accuracy

4) Ambient Temp Ef fect

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.ACCIDNET ENVIRONMENT AND PAST PERFORMANCE Typical errors addressed in Foxboro Test Report No. Q9-6005 indicate per-

- formance within 5% (Point Beach Nuclear Plant assumed + 5% of pressure as more conservative). Note also that errors tend to be negative. Pos it ive errors are assumed since they are more conservative.

Assume Point Beach Nuclear Plant Errot = + 5% of Test Pressure

= + .05 of 2030 psi

= + 102 psi Maximum Error = 11 + 102 = 113 psi (Note that an increasing containment pressure will reduce the pressure output and tend to be more conservative.)

24

l BYPASS RTD'S TE 401A TE 401B TT 401A R/E R/E TT 401B i

dT / T sve Lab n! 405R L/L TM 401BB E/E 100 Span w @I

.I I/I CI T

TM 401t; CI 4

25

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NORMAL AMBIENT ACCURACY BYPASS RTS'S

= + 0.2 F RTD II there are 7 as.nalog devices used to determine

/i T and T,y, - Errors are assumed to be as '

follows :

a) Normal Instrument Accuracy = + .5% of Span

=

][.5%,of100"F

= +

._ .5% F b) Allowed Calibration Tolerance = + .5% of Span

= 7 .5% of 100 F

= T

_5F III RTD Readout for Calibration is

.13 JL at .13R/1 1R/ F = & .08 F IV Computer Accuracy + .125% = + .125 F Repeatable + .125% = + .125 F Re solut ion + .0625% = + .0625 F y{14(.5) + 2(.2)2 + 2(.125)2 +(.0625)2 + (,08)2 , 1,9o 7 Assume Ambient Error of 2.0 F No Accident Environment Error is Assumed l

1 I

26 f

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• 2.1.4 DIVERSE CONTAINMENT ISOLATION No_v ives change position when containment isolation is reset. Manual action, in addition to resetting CI, is required before valves change position.

Also, valves from different systems are not " ganged" together. Repositioning valves in one system will not cause valves in another system to change position.

O 6

27 i

2.1.5.a DEDICATED II, CONTROL PENETRATIONS For each unit, there are two separate air supply paths, and two separate' vent paths. Refer to the Figure 2.1.5.a-1 showing these paths.

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2.1.6s System Integrity for High Radioactivity A. All systems required to cool the core have been included in the l

leak check program; furthermore,the inoperability of an excl6ded

. system will not affect the ability of the included systems to perform their core cooling functions.

B. The concerns of IE Circular No. 79-21 have been addressed,.and compliance with the requirements of the circular will be met, specifically:

i 1. Written procedures are followed when radioactive liquids are transferred. Management audits will be inaugurated by Sept 1980 to assure compliance with such procedures.

2. As built systems having the potential for inadvertent releases j have been in-service tested to detect any design or construction errors.

3. The integrity of systems containing liquid. radioactive material is assured by four hour interval logging of tank levels and sump pump operating times. In service hydro tests are performed on new l or required systems.

i 1

C. Management assurance that the 2.1.6a committments are being met are:

l l 1. Leakage tests are written and scheduled by the plant staff engineers.

The actual test is performed by the shift operators or radiation protection specialists with test results reviewed by the Shift Senior Reactor Operator and an assigned engineer. The completed test becomes part of the plant operating / maintenance history.

2. Leakages found as a result of the leakage tests are. documented on the written leakage test by the pers 7 doing the test and requests for work initiated to resolve the _ let - are written by the person, the Senior Reactor Operator or the assigned engineer.

3. Administrative Control Directive 5ACD 10.1; Control of Radioactivity requires that Superintendent of Radiation Protection conduct radiation surveys, measurements and evaluations for the assessment f

and control of radiation hazards associated with the Prairie Island Plant.

f- 30

.-,e.

, ._,. n.

b I'

., ' These surveys, measurements and evaluations assure that leaks and spills T

i are' detected,' and that good housekeeping practices are implemented.

4

. This ACD along'with the ALARA' program dictates timely correction of leaks.

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.w*-----. ~ . - - - ----e- - + -*- ,--, e,yv -.w--e- ca- -w

2.1.6.b PLANT SHIELDING REVIEW A'n extensive design review of plant radiation levels has been performed using source terms defined in NUREG 0578. The systems evaluated were the High Head

~ Safety Injection System, Residual Heat Removal System, Containment' Spray Sys-tem, Shield Building Ventilation System, Auxiliary Building Special Ventilation System, Primary Coolant System and Containment Sampling Systems. The control room ventilation and chiller areas are in low radiation areas. A close look at the Letdown System indicates that it should not be used in a high activity situation. We will isolate the Letdown System at a predetermined radioactivity level in the event of large fuel failure.

The letdown portion of the CVCS is not needed for accident recovery.

Letdown will not be required to degas the Primary System. The head vent, vent-ing to the PRT and ultimately the containment, will be used to degas the Pri-mary System, should that become necessary.

Letdown is not required to borate the Primary System or to maintain seal in-jection flow. The charging pumps can take suction from either the Boric Acid Storage Tanks (BAST) or the Refueling Water Storage Tank (RWST). The head vent can be used to relieve water inventory from the Primary System, and allow continuation of charging or seal injection.

A RCS cooldown and depressurization can be accomplished using the SGs and the head vent. The head vent can function like the Letdown Systet, only the vent would relieve eventually to the containment instead of the VCT or HUT. The BAST and RWST would supply ample borated water for a cooldown and eventcal RHR System recirculation.

Should the BAST and RWST be empty, because of SIP and/or RHRP injection into the Primary System, it would indicate a large RCS leak. Long term recircu-lation would be initiated, and letdown would not be required.

If, for some unforeseen reason, the RWST and BA5i have been injected into the RCS and the water is now on the containment floor, the head vent can be used to lower the RCS pressure sufficiently to allow SIP injection. The SIPS would get suction from the RHR pumps, which are taking suction from the containment.

RCP seal injection can be maintained for at least eight houra, using the BA that is left in the BAS

• after the tank has been isolated from the SIP suction.

32

2.1.6.b PLANT SHIELDING REVIEW (con't)

. There is at least 850 gallons of 12% BA that can be blended to 1% and injected into the RCP seals. This would allow ample time to batch more BA and cool and depressurize the RCS, eventually going on RHR.

The above systems were physically traced out in the plant to determine high dose areas. Instrument line reference legs in the auxiliary building were assumed to contain pre-accident fluids during the system walkdown. Areas that are considered vital will be shielded. The requirement for < 15 mR/hr in the control room technical support center and operations support center will be met within a few hours after an accident. The integrated dose for an individual in the control room is less than 2 rem for the first 30 days after the accident.

Other areas of the plant considered as vital for short term entry are the sample room, hot lab, counting room, access control / health physics area, radwaste con-trol panel areas, post LOCA hydrogen control area, all safeguard motor control center areas, and all safety related equipment areas outside the auxiliary build-ing. These areas outside the auxiliary building include the cooling water pump areas, diesel generator rooms, the sa.feguard switchgear and battery rooms, and the auxiliary feedwater pump room.

The radiation environmental qualification of the Safety Injection System, Re-sidual Heat Removal System, the Containment Spray System and the Special Venti-lation Systens is continuing. At present time it appears that the equipment is radiation qualified. Required safety related instrumentation in the areas of these systems is either qualified, being qualified, or will be relocated to low dose areas. We are presently evaluating recirculation mode equipment (equipment that receives its supply of highly radioactive water from the con-tainment sump in a post-accident recirculation phase) assuming it is essentially degassed as is the Westinghouse Owners Group recommendation.

Several modifications are being made which will reduce radiation levels in the

- auxiliary building. (1) The installation of the Reactor Head and Pressurizer vent System eliminates the need to use the netdown System. This eliminates the use of the Waste Gas Systems and Non-aerated and Aerated Waste Liquid Systems.

(2) Another modification that will be undertaken is to reroute the residual heat removal (RHR) sump from the waste holdup tanks to the containment. (3) We 33 l

4 Y +

T - - - -

w T

2.1.6.b PLANT SHIELDING REVIEW (con't) will route the Sampling System' drains to the RHR sump. With modifications 1, 2 and 3, all highly radioactive liquids and gaseous leakage will b.e pumped into or remain in the containment.

4. The safety injection (SI) pump area must be shielded with approximately 16 inches of concrete.

5. The opening from the containment spray pump room must be filled in with concrete.

6. The RilR pit covers must be redesigned so that they may be lef t in place except when maintenance or inspection is going on in the pits.

7. The safety injection line for cold leg injection must be rerouted or shielded.

8. The safety injection line for reactor vessel injection must be shielded.

9. The RRR supply to the SI pumps routed in a pit with one foot of concrete above it. This line will need additional shielding.

10. The opening where the RilR lines penetrate the containment spray pump room must be shicided.

11. The 715' level floor area above the RHR lines must be further shielded.

12. The charcoal areas of the shield building vent filters must be shielded.

! 13. The charcoal areas of the Auxiliary Building Special Vent System must be shielded.

14. The Loop B sample line needs shielding in various locations or be rerouted.

15. Misec11aneous shielding modifications in the sample room.

16. The environmental qualification of the RHR sample valve is not clear. The valve will'either be moved, shielded, or replaced. Replacement valves were ordered on 1-3-80.

17. The environmental qualification of the Loop B sample lines in containment is questionable. Replacement valves are on order as of 1-3-80.

34

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4 L 2.1.6.b PLANT' SHIELDING REVIEW (con't) v-

' All modifications not requiring outages will be completed prior to 1-1-81.

4 Outage related work will be completed in the fall of 1980 for Unit'l and Unit 2 winter outage which should start about 1-1-81. If the sample valves j' for Unit 1 do not arrive on time, replacement will be made within one month

}. of receipt of the new valves.

)i Abbreviations l ~ SIP = Safety Injection (High head) pump (s) i RHRP = Residual-Heat Removal (Low head) pump (s) i VCT = Volume Control Tank

) HUT' = CVCS Holdup Tank (s)

. BA = Boric Acid SGS =

Steam Generator (s) t' l

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_. . . - - . , - . - - . , ,. .w_ ._ ...- , ,. - - , _ , . _ _ . . . _ , . . , - _.y_. - -.

2.1.7.b AUXILIARY FEEDWATER FLOW INDICATION TO STEAM CENERATOR Auxiliary feed flows are calibrated annually. Calibration involves introducing a pnuematic dp signal into sensor (transmitter or local indicator) at 5 values including zero and full scale. As found data and as lef t data, if adjustment is necessary, are taken for output of each loop component at each of_the 5 input values.

Steam Generator Level - Wide range are calibrated during each refueling outage (approximately 13 month intervals) using the same method as auxiliary feedwater flow.

Steam Generator Level - Narrow range transmitters are calibrated during each refueling using same method as above. Analog loop components are calibrated annually and functionally tested monthly. Operational check of loops by redundant channel comparison is performed each shift. Steam Generator Level Narrow Range check, test and calibration frequencies are in accordance with Technical Specifications.

36 I

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-2.1.8.a POST ACCIDENT PRIMARY COOLANT SYSTEM AND CONTAINMENT ATMOSPHERE

' SAMPLING SYSTEM Sampling procedures have been written that will enable plant perso'nnel to obtain a containment air sample for radioanalysis or for H analysis.

2 Prairie Island _was equipped with a post accident sampling system for H 2 using a Westinghouse-installed gas analyzer. This gas analyzer utilizes two small air blowers to obtain the sample and force it through a thermo-conductivity type H analyzer. This analyzer will detect H2 fr m 0 to 2

. 100% in three ranges. By utilizing this sample system, a sample can be taken of containment atmosphere using a small bomb with a septum to take a syringe of gas for spectral analysis. A NSP mobile GeLi located at Prairie Island will be used to analyze the gas sample.

Reactor Coolant system samples can be taken in our present sample room.

Samples could be taken from either the RCS loop B sample line or the RHR system sample line. The normal sample room has been modified by installing 4 inches of lead shielding for the large sources in the lab.

Mobile shield panels are also provided for the technicians. A small sample will be drawn and placed directly into a shielded carrier for transport to a hot cell located outside the Auxiliary Building for boron, pH or spectral analysis as may be requested by the Control Room or Technical Support Center.

Oxygen analysis would be done in the sample hood in the sample room. Hydrogen analysis would be done by taking a small pressurized sample in a sample bomb and_ performing the analysis in the hot laboratory.

Boron analysis will be performed in the hot _ cell area using a diluted RCS

~

sample and running a carminic acid and spectrophotometric testing using the method of ~ ASTM D30821975 Edition.

A pH on the undiluted sample could be run'in the hot. cell using a standard combination probe on a pH meter.

37

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

. 2.1. 8. a - continued RCS oxygen analysis will be done in the sample room using an indigo carmine. oxygen test on an undiluted sample.

RCS hydrogen analysis will be done'in the sample room and hot laboratory by,taking a small pressurized sample and' expanding it into a H2sample apparatus. The vented gas space will be sampled and H2analysis performed on a gas partitioner.

Spectral' analysis will be performed on the RCS sample taken to the hot cell area of the turbine building. A portion of that sample will be diluted and taken to the Prairie Island mobile GeLi for analysis. The mobile GeLi ,

system is presently on site, operational,and calibrated.

f A study was made to determine the time and dose for doing the above analysis.

-ltLwas assumed .for ~all analyses except the reactor coolant system oxygen ~

and hydrogen which will be done in the sample room and hot lab, that all other chemistry and radio analyses could not-be done in the present hot lab or counting room. Additional facilities of the hot cell in the turbine build-

'ing and the mobile GeLi system have been pursued as an alternative. Below is a table of estimates of the' time and exposure for performing the indicated tasks based on performing the procedures:

e TASK _ _ Time- Whole Body Dose ExtremityDose obtain sample- 20 minutes- 340 mrem 2500 mrem

. Dilute sample &: perform

-spectral analysis 40 minutes 1 mrem from sample 250 mrem-Perform boron analysis 90. minutes 21 mrem 5000 mrem (Boron analysis via carminic acid method requires cooling time and develop-

! ' ment-time.: Both of these times may be reduced with a slight. loss of accuracy,

- to a time .of. about one hour)

Perform pH Sfminutes- Few mrem- 20 mrem Perform H2 analysis 25 minutes, 1080 mrem 4800 mrem L

' Perform 0 analysis 2

20 minutes 570 mrem 2600 mrem

~

l l

38 >

, , - - . _. _, . . . - . _ _ . - - - .. a _.u--..---- _ _ - _ _ . _ _ _- -- . . _ _ .

'O 9 2.1.8.a continued A chloride test can be done using a diluted sample and using the nitric acid-silver ntrate test with a turbidimeter. However, the analysis will have interference from the iodine. There presently seems to be no need for the data. We feel the analysis is not necessary.

There are several modifications proposed for the sampling system. We intend to purchase and install process equipment for RCS H2 and 0 2 measurements. More shielding of RCS sample lines will be added in the sample room to reduce exposures. Also, a manipulator system for obtaining grab samples is being purchased and installed. Installation of this equipment is not outage related and will be completed before 1-1-81.

See section 2.1.6.b for a discussion of the sampling system valves.

39

o .

2.1.8b Increased Range of Radiation Monitors

~

The in-containment high radiation monitors (up to 107 R/hr photon radiation only) will be purchased and installed in Unit 1 during the Unit 1 fall 1980 refueling outage and in Unit 2 during its refueling outage starting about 1-1-81. We presently have placed an order to General Atomics for their high range radiation monitor which appears to meet ANSI N 320-1978. Electrical penetrations for these monitors will be installed during the refueling outages.

Post accident ventilation of the Auxiliary' Building is made via the Auxiliary Building special ventilation system to the Shield Building vent. Each Shield Building vent is equipped with one low range radiation monitor. An additional sample line has been routed to the Turbine Building and equipped with a particulate and silver zeolite filter and also a sample chamber. The sample blower is powered from a safeguards power supply. Interim procedures define how the release activity can be determined using portable instruments and communciation with the Control Room. A Victoreen area monitor which' will be calibrated to read out the release rate in the Control Room has been purchased. It will be installed as soon as practical I

after arrival (prior to 1-1-81) which will allow the Control Room full scale release rate measurements to 104 p ci/cc. No direct containment releases are made post accident. This system will be on scale with releases at abour 20,000 curies per second per stack.

l Both stacks are in service during the accidents. The air ejector l discharge will be rerouted to the shield building stack prior to 1-1-81. All other ventilation discharges will be stopped following an-accident. v The shield building vent stack monitor is located in the Turbine Building to reduce exposure for personnel removing sampling media. Silver zeolite filters will be used to sample for iodines to reduce'the occupational source posed by noble gases collecting on. charcoal filter-media. Portable shields are being constructed that will be available if the sample media is greater than 10 mR/hr.

40

e e Sample media will be counted on the Prairie Island mobile Geli system. The mobile Geli is calibrated for high activity samples.

The use of silver zeolite should reduce the rad levels.on the sample media. The mobile Geli system can run on plant power or it is equipped with a backup generator system. If the normal location of the mobile GeLi is in a high background area at the time, it can be easily moved to a low-background area. The mobile Geli receives a weekly calibration check with a NBS traceable standard.

1 The interim procedures for noble gas release are written to use any of the following equipment calibrated to Xe-133 equivalent:

A. Eberline Teletector Total Model 6112

1. Range -0.1 mr/hr to 1000 r/hr

2. Sensitivity -0.1 mr/hr

3. Energy Dependence i 20%

4. Calibration Frequency - semiannually with sources traceable to NBS B. Eberline RM-14 with HP-210 Probe (For Low range only)

1. Range-O to 50,000 cpm

2. Sensitivity 50 cpm

3. Calibration Frequency J Semiannually with electronic and source calibration traceable to NBS C. Eberline PIC-6A

1. Range, 1 mr/hr to 1000 R/HR

2. Sensitivity 1 mr/hr

3. Enprgy Dependence,

• 10% from 60 KEV to 1.3 MEV

4. Calibration frequency - Semiannually with sources traceable to NBS D. Victorcen Cutie Pie Model 740

1. Range 1 mr/hr to 25 R/hr 2.

Sensitivity - 1 mr/hr

3. Accuracy

• 10%

4. Energy Dependence

• 10%.from 7 Kev to'2 Mev 5 .' Calibration Frequency - Semiannually with sources traceable to NBS Utilizing installed equipment and the above equipment covers the range of~ releases from normal levels to greater than 10,000 C1/sec Xe-133 equivalent. The location of the monitoring points are a i 41

l very short distance from the operations support center and frequent readings can easily be given to the control room or Technical Support Center using the phone system or a hand radio.

The permanent installed meter will be a Victoreen area monitor which reads out from 0.1 mR/hr to 107 mR/hr. It will be calibrated in p ci/cc Xe-133 equivalent. The range will be approximately 10-2 9 ci/cc to greater than 104 pei/cc. The monitor is located in the turbine building in a shielded area to assure background is low.

' The samples are taken from isokinetic nozzles in the Shield Building Vent line. The monitors will be powe ed from Safeguards Electrical systems. The monitor will indicate in the control room and to the plant computer which makes it available to the TSC.

A system will be installed to determine the activity in the main steam lines. We are still evaluating systems at this point and have contracted with Fluor Power Services to develop this system.

The system will take a main steam sample, cool it, and monitor it for radioactivity. Monitors will indicate the status of the safety relief valves and power operated reliefs to obtain a flowrate.

i 42

Sample media will be counted on the Prairie Island mobile Geli system. .The mobile Geli is calibrated for high activity samples.

The use of silver zeolite should reduce the rad levels =on the

-sample media. The mobile Geli system can run on plant power or it is equipped with a backup generator system. If the normal location of the mobile GeLi is in a high background area at the time, it can be easily moved to a low-background area. The mobile Geli receives a weekly calibration check with a NBS traceable standard.

The interim procedures for noble gas release are written to use any of the following equipment calibrated to Xe-133 equivalent:

A. Eberline Teletector Total Model 6112

> 1. Range -0.1 mr/hr to 1000 r/hr

2. Sensitivity -0.1 mr/hr

3. Energy Dependence i 20%

4. Calibration Frequency - semiannually with sources traceable to NBS B. Eberline RM-14 with HP-210 Probe (For Low range only)

1. Range 0 to 50,000 cpm

2. Sensitivity - 50 cpm

3. Calibration Frequency - Semiannually with electronic and source calibration traceable to NBS C. Eberline PIC-6A

1. Range, 1 mr/hr to 1000 R/HR

2. Sensitivity 1 mr/hr

3. Energy Dependence,

• 10% from 60 KEV to 1.3 MEV

4. Calibration frequency - Semiannually with sources traceable to NBS D. Victoreen Cutie Pie Model 740

1. Range 1 mr/hr to 25 R/hr

2. Sensitivity - 1 mr/hr

3. Accuracy i 10%

4. Energy Dependence i 10% from 7 Kev to 2 Mev

5. Calibration Frequency - Semiannually with sources traceable to NBS Utilizing installed equipment and the above equipment covers the range of releases from normal levels to greater than 10,000 Ci/sec Xe-133 equivalent. The location of the monitoring points are a 41 l 1

very short distance from the operations support center and frequent readings can easily be given to the control room or Technical Support Center using the phone system or a hand radio.

The permanent installed meter will be a Victoreen area monitor which reads out from 0.1 mR/hr to 107 mR/hr. It will be calibrated in p ci/cc Xe-133 equivalent. The range will be approximately 10-2 y ci/cc to greater than 104 pei/cc. The monitor is located in the turbine building in a shielded area to assure background is low.

The samples are taken from isokinetic nozzles in the Shield Building Vent line. The monitors will be powered from Safeguards Electrical systems. The monitor will indicate in the control room and to the plant computer which makes it available to the TSC.

A system will be installed to determine the activity in the main steam lines. We are still evaluating systems at this point and have contracted with Fluor Power Services to develop this system.

The system will take a main steam sample, cool it, and monitor it for radioactivity. Monitors will indicate the status of the safety relief valves and power operated reliefs to obtain a flowrate.

42

e e 2.1.8.c IMPROVED IN-PLANT RADIO IODINE MONITORING 1

The~ Technical Support Center and the control roem will have iodine continuous air monitors available for post-accident air monitoring.

Silver zeolite cartridges will be used with a single channel analyzer looking for the iodina 131 peak at 365 kev in these CAM's.

, A mobile GeLi System is located at Prairie Island. (The system is operable and calibrated at this time.) This system enables the plant technicians to take particulate and lodine samples at desired locations in the plant.

These samples can then be accurately analyzed. Operating procedures and

, preliminary training have been completed on the mobile GeLi System.

e

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I 1

i 43

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, , .4- . - - . . - -c. . ., . - .- , - --

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-'2.1.9 ' REACTOR VESSEL HEAD AND PRESSURIZER VENTIJr ,

The existing 1"'line f rom the reactor head has been analyzed for a rupture.

Page 4.3-13 of the Prairie Island FSAR states, " Pipe rupture analyses were

_ performed on all high pressure piping (including lines with diameters less than 3/4"). The analyses establish that the containment vessel and all {

essential equipment within.and without the containment (system and equip-ment defined as Class I in App. B of the FSAR) are adequately protected against the effects of potential pipe ruptures".

There will be a pressure indication sensing the line between the isolation valves to detect if there is leakage past the high pressure valves.

All valves in the RCS Vent System are of a fail-closed design, double iso- ,

1 lation is also'provided, and are powered by separate electrical trains.

We will take all necessary steps to minimize inadvertent opening of the vent valves.

4

?

a 4

L l

l-44

-L-'- J - aLs - - w, - m. 4 -

s -s l

2 2.2.1.a SHIFT SUPERVISOR RESPONSIBILITY The following Administrative Control Directives and Section Work Instructions define the Shif t Supervisor's responsibilities, duties, requirements,-training and authority:

1ACD3.2 Corporate Level Plant Operations 3ACD3.2 Power Production Level Plant Operations 3ACD3.5 Power Production Level Training j SACO3.1 Plant Level Plant Organization 5ACD3.ll Plant Level Plant Training Program SWI-0-2 Shift Organization, Operation & Turnover These directives and instructions were modified to reflect the requirements of NUREG-0578 and liarold Denton's letter of clarification. SWI-0-2 and 5ACD3.1 require the Shif t Supervisor to remain in the control room at all times during accident situations unless properly relieved. Persons authorized to relieve the Shif t Supervisor are specified.

i I

f a

45

2.2.1.b SHIFT TECHNICAL SUPPORT

.- Prairie Island presently has eighteen Shift Technical Advisors who work

. on 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period providing continuous on-site coverage. Duties, requirements and responsibilities are defined in the Section Work Instruction SWI-PERP-15.

During accident conditions the STA acts as an advisor-to the Shift Supervisor and has no other functions or responsibilities. He is available to the Shift Supervisor in the control room within 10 minutes.

The STA's are all degreed individuals,all but one possessing engineering degrees.

All presently have SRO licenses or have had SRO licenses on the Prairie Island Plant. We plan to expand the group by adding about 6 new STAS for the second half of 1980. These individuals have engineering degrees and will have received training in operations and accident assessment. This training will include experience at a simulator. All the STAS will receive additional training on transient analysis, accident assessment, thermo-dynamics and other selected subjects pertinent to STA requirements. This training is planned for early this year.

Operational assessment functions are carried out by the plant Engineering Staff, the plant Operations Committee and the Safety Audit Committee. The focal point of assessment centers around the Staff Engineer. His responsf5111ty includes assignment and review of investigative reports for requirements such as Reportable Occurrences, Significant Operating Events, NRC letters, Bulletins, Circulars, and Information Notices. He assesses the importance of other operational material from other similar plants supplied by various groups such as Edison Electric Institute. This individual is an engineer, possesses an NRC Senior Operator License and is a STA.

The Engineering Staff, the Plant Operations Committee and the Safety Audit Committec

. review all Reportable Occurrences- Investigative Reports and Significant Operating Event Investigative Reports other pertinent reports are reviewed by selected members l-I 46

a .

2. 2.1.b (continued) of the Engineering Staff for assessment. Recommendations and Corrective Actions are reviewed and there is a system established to assure follow through.

47

e .

2.2.1.c SHIFT' & RELIEF TURNOVER PROCEDURE A checklist is presently in use for shift turnover, which defines shift change status information between Shif t Supervisor, Lead Plant Equipment and Reactor Operator and Plant Equipment and Reactor Operator. Information includes significant equipment / system not availabic for service, signifi-cant equipment repaired and made available for use during the shif t, out-standing surveillance procedures, significant work requests, operational plans for the coming shif t, and new operational and administrative pro-cedures. Important Technical Specification limits and plant parameters are indicated on control room annunicators.

i' Other plant sections such as maintenance and instrument technicians normally do not work shifts. All work is controlled by the control room so that critical system line up and status is the control room responsibility.

t The turnover procedure is reviewed by the Shif t Technical Advisor each shif t.  !

The evaluation of the effectiveness of the turnover procedure is a line function and like other procedures of this nature is assessed by the operations

.; organization.

48

4 'O 2.2.2.b ONSITE TECHNICAL SUPPORT CENTER 1/1/80 Requirements The Technical Support Center (TSC) is established in the

" Engineering Conference Area" of the plant administrative office. As built plant records that include general arrangement drawings, piping and instrument diagrams, piping system isometrics, electrical schematics, wire and cable lists and single line electrical designs are available to this area. The Technical Support Center is located across the Turbine Building from the Units 1 and 2 main control room.

The procedure for providing the engineering / management support and staf fing of the TSC is provided as a Temporary Memo to the existing approved Emergency Plan Operation Manual, Section F3.

The TSC will be activated when a " Site Emergency" as defined in our Emergency Plan is declared or whenever it is deemed necessary by plant management. If activation of the TSC occurs during thc normal working day, the onsite members of the Operations Committee will report directly to the center. If activation occurs during off-duty hours, the Duty Engineer will be contacted and will contact the appropriate members of the management staff depending on the type of expertise needed.

Communications between the Technical Support Center, the near site Emergency Operations Center, the Control Room and the Nuclear Regulatory Commission have been established. The Technical Support Center and the near site Emergency Operations Center are in the same location.

A dedicated communication link between the TSC and the NRC has been installed by the telephone company. Communications between the TSC and the Control Room will utilize the plant telephone system. ~ Backup communication will be provided by the plant Sound Powered communications system. Alternate backup communi-cations can be supplied by plant portable radios. The NRC " hot line" phone extension has been installed in the TSC.

A portable continuous air sample airborne monitor and a portable whole body monitor have been installed in the TSC. Upon reaching high radiation levels in the TSC, selected management and engineering support personnel will evacuate to the control room. Other non-essential personnel will be evacuated.

49

o -

A Unit 1 and Unit 2 Computer." Engineers Console" is located in the TSC. This console has access to all data points accessed by the plant processor via a keyboard and output typer. In addition, a " slave" CRT is provided to display a dedicated group and alarm group of parameters as selected by the control room operator. A notebook containing flow diagrams, computer addresses, instrument numbers, control room control board identification numbers and ranges for the parameters needed for accident assessment in the TSC will be provided in the TSC in 1980.

Upon evacuation from the TSC, selected individuals would be dispatched to the control room. By using the information in the

" notebook" described in E, accident assessment could be performed f rom the parameters displayed on the control board.

1/1/81 Requirements Plans are being made to upgrade the TSC by 1/1/81 as follows:

a. Physical Size and Staffing The TSC will be sized to accommodate approximately 25 people. A minimum of 600 square feet will be available in the TSC for equipment and personnel. The main functional areas in the TSC will be: recording and display panels, plant process computer engineers consoles, conference area and communication cubicles.

The TSC is located in close proximity to the normal plant administrative area such that records of as-built conditions and drawings of structures, systems and components will be readily available. Also, normal administrative functions will be available as needed. Because the use of these functions will be limited and protective measures could be applied to their use, the administrative functions will not be included in the habitability envelope of the TSC.

The maximum staffing level for the TSC is anticipated to be-as follows:

a. 5 NRC personnel including a site inspector

b. 3 Westinghouse members of the " site response team"

c. 1 Corporate Management Representative

d. I communications person 50

4 .

e. 1 administrative specialist

f. 11 members of plant management (Operations Committee)

g. The remaining personnel would consist of technical specialists from the plant staff or outside consultants as appropriate for the occurrence.

b. Building The area being considered for the TSC is presently designated as the Administration Building Annex. The structure is located on the same elevation as the Control Room, approximately 100 feet away across the Turbine Room floor. Easy access from the TSC to the Control Room is, therefore, assured.

The area under consideration is large enough to house 25 persons, informational displays, radiation monitoring equipment and the necessary technical data.

The structure is constructed of concrete blocks and precast concrete panels. This structure can be modified to afford protection f rom direct radiation as well as airborne. A pressurizing-type ventilation system will be added which will include particulate and charcoal filtering.

It is anticipated that the backup electrical power supply to the HVAC, lighting and radiation monitoring equipment will utilize the diesel generator which provides the backup for the plant security system. Since sufficient reserve capacity is not available on the present security system diesel generator, replacement with a larger capacity diesel generator is foreseen.

c. Communications A minimum of three direct telephone lines to the Corporate Office Telephone Exchange will be installed. These lines are part of Northern States Power's microwave communications system and do not depend upon the plant telephone system or the local telephone system. In addition, a multi-channel radio phone will be installed in the TSC. The radio will have the ability to communicate with the following:

1. Sheriff's Department

2. NSP Dispatcher and State Emergency Operations Center

3. Portable Radios (Plant)

4. Mobile Radios (Trucks, etc. )

51

. d. Information Display and Storage The engineers computer console and a data display system will be utilized to provide access to the parameters needed by the TSC staf f to assess the consequences of and supply engineering support to the control room operating staff following an accident or severe transient on the Nuclear Steam Supply Systen.

Most of thia information is available to the engineer's console at this time. By 1981, all vital information will . be available to the Technical Support Center via the engineer's console and a data acquisition and display system.

The acquisition'and display system will provide the capability for historical and current status display of the following key parameters.

NOTE: For historical purposes, the initiating event will be considered time zero.

TECHNICAL SUPPORT. CENTER KEY PARAMETERS Parameters Notes I. Reactor Coolant System A. Temperatur'e RTD's and/or

<. thermocouples B. Pressure Wide range C.- Pressurizer Level II. Secondary A. Main Steam Pressure Each Stm Gen B. Steam Generator Level Wide range III. ECCS~ System A. S.I.-Flow

'IV. Feedwater and Makeup System A. Aux FW Flow Each Stm Gen V. Containment A.- Pressure Wide range B. Level Wide range 52 ,

a e Key Parameters (con't)

Parameter Notes VI. In-Plant Radiological A. Reactor Coolant System Data from manual sample B. Containment High range C. Effluent Activity (Noble Gas) Shield Building Stacks (2)

VII. Offsite Radiological A. Meteorological

1. Wind Speed

2. Wind Direction B. Offsite Radiation Levels Data from manual sample

e. Instrumentation Power Supply The power supply for the computer console and the data display system will be continuous once the TSC is activated. Loss of offsite power or switching transients will not result in a loss of key parameter historical data.

f. Data Transmission Northern States Power Company is participating in a Westinghouse Owner's Group Ad Hoc Instrument Systems Committee that has in its scope the investigation of data transmission methods.

This Committee is addressing many long-term concerns in the area of data acquisition, transmission and display. By taking i an integrated aproach to those many concerns, it is anticipated that a more systematic solution can be arrived at to address  ;

related long-term concerns. ,

l 53 i

3 o c-2.2.2.c OPERATIONAL SUPPORT CENTER The interim Operations Support Center is located in the Records Room next to the Control Room. The room is quite large and can accommodate about 25 people.

Presently we have one telephone in the Operational Support Center. Another has been ordered and will be installed soon.

Communications will be upgraded by 1-1-81 to include an intercom system be-tween the Operational Support Center, the Technical Support Center and the Control Room.

54