ML20083N900

Rev 2 to Procedure RTP-34, Contamination Radiation Monitoring - Post-Accident

ML20083N900
Person / Time
Site:	FitzPatrick
Issue date:	03/23/1984
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20083N885	List: ML20083N882 ML20083N896 ML20083N900
References
RTP-34, NUDOCS 8404190325
Download: ML20083N900 (8)

Text

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ATTAClelENT. C. .. n ?

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..l JAMES A, FIT 2FATRICK NUCLEAR POWER PLANT'-

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)!t RADIOLOGICAL AND ENVIRONMENTAL SERVICES DEPARTMENT ',

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PROCEDURE No.

RIP-34 1 1

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3 1 TITLE: CONTAINMENT RADIATION MONITORING - POST ACCIDENT

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[ PORC Review No./Date Meeting No. [4 - o#-d Date g/4 /M i .

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i Approved By: Y /X Resident Man &gelr

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'f Approved By: o j Radiological Mrfd Envi mentsT,__,___

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Rev. No. 2 -

Date ,03/84

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..f RTP-34 l

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CONTAINMENT RADIATION MONITORING - POST ACCIDENT

• 1.0 PURPOSE The purpose of this procedure is to describe the containment

)f high range radiation monitors, operation and use.

f 2.0 REQUIREMENTS AND ACCEPTANCE CRITERIA

f. 2.1 Technical Specifications None 2.2 Plant t

2.2.1 Containment high range radiation monitors should be oper-able during operation of the plant.

2.2.2 During each refueling outage, the containment high range radiation monitors should be source checked.

[ 3.0 SPECIAL EQUIPMENT t

3.1 Containment High Range Radiation Monitors Two radiation monitors (27-RM-104A and B) provide gamma ra-diation indication of the primary containment. These moni-L tors were designed to meet the requirements of NUREG-0578 and NUREG-0737.

The General Atomics' Model RD-23 gamma radiation detectors are widely separated and located in the large volume portion of the drywell. Detector "A" (27-RE-104A) is located on the i south side of the drywell at elevation 290 feet four inches,

' above 10-MOV-18 near the ladder to the 292 foot elevation.

Detector "B" (27-RE-104B) is located near the east recircu-lation discharge line at elevation 288 . feet four inches, above the east unit cooler.

Each detector has an internal U-234 " bug" source to provide a minimum detector current corresponding to 1 R/hr.

The' General Atomics Model RP-2C readout modules are located in the main control room on process radiation monitor panel (09010)'. The readout log ratemeter displays the detector indication over' a range of 1E+1 to 1E+8 R/hr. Front panel' controls include three alarm lights / reset buttons, and ' a spring loaded, three-position function switch for normal operation, trip level adjustments, and internal check signal-actuation. In the CHECK position, an electric current is Rev. No. 2 Date 03/84 Page 1 0

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i generated which produces a mid scale indication. Test points i and adjustment potentiometers are accessible from the front

[ p panel.

p Each containment radiation monitor has " FAIL", " ALERT", and

"HI-HI" annunciator alarms located on the main control room panel (09-4). Computer alarms F545, F546, F547 for 27-RE-104A are FAIL, HI, and HI-HI alarms respectively. Computer 1

alarms A564, A565, A566 for 27-RE-104B are FAIL, HI, and a

H1-HI' alarms respectively. Analog computer points E026 and

! E027 are available for "A" and "B" monitors respectively.

3.2 General Atomics RT-11 Calibrator L The General Atomics RT-ll Calibrator, weighing 56 pounds, l

consists of a Cs-137 gamma source in a removable housing and a shielded carrier. The source is positioned next to the

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RD-23 gamma detectors to verify that the detector responses to radiation. An activity of about 100 mci is distributed 1 -

in a line which allows this source to produce an on scale

i. reading of approximately 10 R/hr.

I When not in use, the RT-ll is contained in the portable r shielded holder which reduces surface exposure rates to

'E reasonable levels. The source is basically a lead brick-f with attenuators, and must be handled carefully.

V l 4.0 PROCEDURE r

j . 4 .1 Containment Exposure Rate Evaluation i

1 4.1.1 A document entitled " Procedures for ' the Determination of the Extent of~ Core' Damage Under Accident Conditions" iden-

[ tified by General Electric as NEDO-22215 provides a method vc to estimate core damage based on . containment radiation levels. This procedure uses the'above document.to inter-pret containment radiation monitor data under ' accident 7

b conditions.

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4.1.2 Obtain the containment radiation monitor reading from the control room in rem /hr.for each available monitor and re-cord data on Figure 1.

4.1.3 Determine.the elapsed time from plant shutdown to the mon-

-itoring reading ~in hours as follows:

a. Record time of reading'on Figure 1.

b. Record time of shutdown on Figure.l. ,

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Calculate hours since shutdown-'and record on Figure 1.

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g T 4.1.4 Normalize the radiation monitor to the reference plant

[ using the following formula:

[f (rem) normalized = rem (1,670+P) (V+237, 450) (6*D)

Where f f P = FitzPatrick reactor power, 2,436 V = Total containment free volume, 267,000 f _ D = Distance of detector to shield wall; i

Approximately five feet for 27-RE-1044 Approximately' ten feet for 27-RE-104B

a. For 27-RE-104A (rem) normalized = ren x 0.925

b. For 27-RE-104B.

(rem) nornalized = rem x 0.463

c. Record the normalized radiation data on Figure 1.

'4.1.5 Using Figure- 2, -

determine the fuel inventory release. *

, Record the best estimate for each detector on Figure ~1.

4.2 Detector Source' Check

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4 '. 2.1 This 'section of the procedur9 shall be coordinated with the Instrument and Control Department surveillance test'of this system. Each detector is exposed to a radiation source by.the Radiological and Environmental Services De-partment.and the Instrument and Control -Department veri-fies that the mo".itor is responding to radiation.

a. Assure that the Instrument and. Control Department is fully prepared to obtain the necessary' data for veri-fication'of monitor response to radiation.

b. Two technicians transport the- radiation source. and holder to the 292 fcot platform in the drywell. and

~c lear the' area around and below the-detector'to mini-mize' radiation exposure to other workers.

Rev.'No. 2 Date: 03/84 L Page 3

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c. One technician removes the source from the holder and transports it to the appropriate detector.

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d. The other technician leaves the drywell and notifies

, the control room that the source is being placed on the detector,

e. The source is to be held with the source line parallel 4 to and touching the cylindrical ptrt of the detector f~ housing.

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[ f. The control room personnel obtain necessary data as I; rapidly as possible, f g. The control room personnel notify the drywell techni-cians immediately when the data has been collected,

h. The source is then removed and placed back into the s

shielded carrier.

i. If the second detector is to be checked, leave the

- source secured in the drywell. Leave the drywell un-til the control room is fully set up and then repeat the check procedure for-the other detector..

j. When complete, notify . an. RES Supervisor of the re-sults.

5.0 REFERENCES

5.1 Procedures'for the Determination of the Extent of Core" Dam-age Under. Accident Conditions, NEDO-22215, General Electric.

5.2 J.A.F. Modification F.I.- 80-16' Containment High Range Radia-

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tion Monitors.

5.3 High Range Gamma Radiation' Monitoring System Operation tand 7 Maintenance Manual No. E-115-876, General Atomics Company, 1981~.

, '5.4 RT-ll Calibrator Descripti'n:

o and -Operating Instructions No.'E-255-1040,. General Atomics, October 1981.

6.0 ATTACHMENTS

.6.1 Figure'l',; Containment Radiation Monitor Data / Work Sheet 6.2. Figure.2, Approximate _ Source and Damage Estimate

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[ CONTAINMENT RADIATION MONITOR DATA / WORK SHEET h

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Date:

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I A. Radiation Monitor Data 17-RE-104A Rem: Date: Hour:

( 17-RE-104B Rem: Date: Hour:

B. Elapsed Time (hours)

I Shutdown Date: Time: k,

, Hours Since Data Obtained: Hours C. Normalized Radiation Data l 17-RE-104A rem x 0.925 = --

reta (normalized) 17-RE-104B rem x 0.463 = rem (normalized)

D. Best. Estimate of Fuel Inventory Release from Figure 2 17-RM-104A 17-RM-104B l

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APPROXUIATE SOURCE AND DAMAGE ESTIMATE .I l ___ . _ _ _ _ . ---- i j percent of Puel 2nventory Airborne in the Contaisement- - - -  ;

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, { 1004 Fuel Inventory = 1006 Noble Gases j l b,

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Inventery i Approximate sor.rce and Damage Estimate Released 100, 2006 T1D-14I44, 2004 fuel damage, potaatial i core melt.

E 50. 30t T2D noble gases, 2MI aource.

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3. 36 TID, 2004 IfA53-1400 gap activity, major elad failure. l

1. 14 T1D, 19% NBC gay, ataa. 10% alad fallare.

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Rev. No. 2 Date 03/84 Page 6

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1 NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT JPN-8 4- 2 3 k

ATTACHMENT D 4- e e

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.m m ..m.o 3 GENERAL h ELECTRIC j

j NUCLEAR POWER SYSTEMS DMSION GENERAL ELECmc COWANY e 175 CURTNER AVENUE o SAN JOSE.CAttC*NtA 95195 MC 682, (408) 925-5040 NFN-006-84 FRH-003-84 1

l January 18, 1984 3 U. S. Nuclear Regulatory Comission Office of Nuclear Reactor Regulation Division of Licensing Washington, D.C. 20555

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f Attention: Darrell G. Eisenhut, Director Gentleman:

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SUBJECT:

ACCURACY OF DISSOLVED GAS: MEASUREMENT FOR GE POST-ACCIDEN"I

! SAMPLING SYSTEM (PASS)

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Reference:

Meeting, between NRC management (R. Vollmer, et.a1.) and

. GE (J. Quirk, et.al.) dated December 12, 1983 on the subject of need for dissolved gas measurement in BWR.

The purpose of this letter is to docyment the agreements reached between the NRC Staff and GE at the reference' meeting and to justify the adequacy of the currently installed PASS for measurement of dissolved gas.

l i At the reference meeting, GE provided a technical justification (Attach-

. ' ment 1) supporting our contention that post-accident monitoring for dissolved gases in a BWR is not necessary. In addition, GE is providing the expected hydrogen partitioning .between primary water, reactor vessel steam dome and containment (Attachment 2). These calculations assume the '

reactor to be at high pressure,.thich maximizes the possible dissolved -

hydrogen. They also assume thet a sample is taken very soon after a postulated significant metal-water reaction event, before continued safety / relief valve actuation complete 1;r flushes the primary system of hydrogen.

Based on these results, GE continues to feel that such measurements are

=i not needed in order to address NRC's concerns of estimation of degree of core damage and assessment of coolant corrosion potential. Desp!to this 1 position, GE has supplied a dissolved gas measurement capability as part

' o of the PASS design.

ii The GE PASS quantifies the concentration of total dissolved gas by

- allowing the expansion of the dissolved gas from a known volume of liquid 3

into a partially evacuated gas collection region and measuring the

.g pressure rise in that gas collection regfon. The accuracy of this measurement for the GE PASS has been. determined by GE to be at least 1505

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l GENERAL U. S. Nuclear Regulatory Commissionh ELECTRIC Page~2 3

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for, dissolved gas concentrations between 25 cc/kg and 50 cc/kg and at least 130% for dissolved gas concentrations greater than 50 cc/kg.

j GE feels that these accuracies are sufficient for dissolved gas measure-j ment in the BWR for the following reasons. At low pressure following an 3 accident, the gases evohed from the core will be essentially all released

] to the containment atmosphere, and the dissolved gas concentration for j these conditions will not be useful. Even under postulated high pressure 1- conditions following an accident, Attachment 2 shows that, for a postulated 1

' dissolved gas concentration of 50 cc/kg: (1) the weight of hydrogen in the reactor pressure vessel is at most 5% of the total hydrogen generated i by the metal-water reaction due to discharge of the safety / relief valves to maintain system pressure and (2) the dissolved hydrogen is that associated with the reaction of only 0.5% of the cladding with water.

Consequently, even if the dissolved gas contribution is totally ignored 3

under postulated high pressure conditions and a dissolved gas concentration 3 of 50 cc/kg, the absolute error in the total hydrogen release for estimation g of core damage would be at most 0.5% metal-water reaction.

) GE analyses have demonstrated that, for high pressure conditions ($1050

, psig) and total dissolved gas concentrations of greater than 25 cc/kg, the dissolved gas composition is essentially all hydrogen. As noted previously, at low pressure conditions, the dissolved gas concentration L would be too small to be useful as a core damage indicator. The NRC staff position as given at the reference meeting was that monitoring for dissolved hydrogen was sufficient to preclude the need for monitoring of dissolved oxygen. .

i GE will be providing procedures to all of its PASS users for measurement b

of dissolved gas.

i ll A prompt response to this letter would be appreciated. If you have any 4, questions, please call J. F. Quirk at (408) 925-2606 or F. R. Hayes at

[ (408) 925-2140 of my staff.

h Sincerely, i

G G. Sherwood, Manager Nuclear Safety and Licensing Operation GGS:csc/112143 cc: R. Johnston (NRC) 7. Sullivan (NRC) F. Witt (NRC)

R. Vc11mer (NRC)

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t NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT JPN-84-23

,. ATTACHMENT E t*4.,

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ATTACIIMENT E I

, The following table of " Integrated Dose Rates" was previously

- submitted in the Authority's phase 1 responses to NUREG 0737 Item II.B.3 (Reference 2) . This revised table is being submitted to incorporate calculated background doses with Reg.

Guide 1.4 source terms one ( 11 hour1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> after an accident. Computer codes developed for response to NUREG 0578 Item 2.1.6.b were used as the basis for these results. A 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> background dose was applied for the isotopic analysis due to the approximate time sequence for obtaining and transporting the sample.

As stated in Reference 2, the times shown are estimated times the whole body or extremity is being exposed to its respective dose and not the actual time required for the particular task.

Sample dose rates were obtained from Reference 3, with plant unique correction factors applied for JAF as follows:

Coolant Inventory = 0.927 X Reference Plant Drywell and Torus Air Volume r- 0.839 X Reference Plant The FitzPatrick Plant, with these revisions, still complies with the personnel radiation exposure limits of 5 rem whole body and 75 rem to the extremities.

Note: Some of the times have been increased to allow for unanticipated equipment malfunctions.

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,g ISTEGRATED DOSE RATES (1 hr. after accident) Rev. 1 0

. c Notes: E = Extremity doses calculated at 10 cm.

W = Whole body doses calculated at 60 cm.

I 1. Liquid Sample (small Time Background Sample Dose Integrated Dose volume) (min) (mr/hr) (mr/hr) (mr) t (W) Sample Sink pre-op 15 103 -

26 (W) obtain sample 45 103 93 147 n

(E) handle sample .5 103 1300 11.7 L

(W) transport cask 10 100 5.7 17.6 (W) cample preparation 6 640 220 86 I (E) sample preparation 4 640 .7920 571 n'

(W) cample analysis 8.5 (borod 640 44 97.5 (E) sample analysis 2.0(borod 640 1580 74-(Z) sample analysis 20 (isotopic) 400* --

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2. Gas Sample (Drywell atm.)

(E) Sample sink pre-op 15 103 26 (W) obtain sample 30 103 302 203 (Z) handle sample 1 103 344 7.5 (W) transport cask 10 100. 45.3 24.2 (E) transport cask 10 100 20,100 3366.7 (E) carple preparation 6 640 520 116 (E) cample preparation 4 640 20,100 1383 (Z) cample analysis 30 (!!ydrogen) 640 18.5 329.3 (E) cample analysis 1.5 (llydroged 640 663 32.6 (W) cample analysis 20 (isotopic) 400** --

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) Totn1 integrated whole body dose for liquid sample = 507.1 mr.

T:t:1 integrated whole body dose for gas sample = 839.0 mr.

T:tal integrated whole body dose for samples = 1346,.1 mr.

Total integrated extremity dose for liquid samples = 6'56'.7 mr.

Total integrated extremity dose for gas samples 4782.3 mr.

Total integrated extremity dose for samples = 5439.,0 mr.

Eassumed 2 hrs. after accident