05000255/LER-1980-021, Forwards Corrected App C, Calculations of Radiological Effects as Results of Hypothetical Accident W/CV-3030 Open, to 800820 LER 80-021,Revision 1.Corrected Apps E & F Also Encl

Forwards Corrected App C, Calculations of Radiological Effects as Results of Hypothetical Accident W/CV-3030 Open, to 800820 LER 80-021,Revision 1.Corrected Apps E & F Also Encl

ML18046A195
Person / Time
Site:	Palisades
Issue date:	11/24/1980
From:	Hoffman D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
References
NUDOCS 8012020580
Download: ML18046A195 (17)

LER-1980-021, Forwards Corrected App C, Calculations of Radiological Effects as Results of Hypothetical Accident W/CV-3030 Open, to 800820 LER 80-021,Revision 1.Corrected Apps E & F Also Encl

Event date:
Report date:
2551980021R00 - NRC Website

text

s consumers Power company General Offices: 212 West Michigan Avenue, Jackson, Mlchl911n 49201 * (517) 788-0650 November 24, 1980 Mr James G Keppler Office of Inspection and Enforcement Region III U S Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, IL 60137 DOCKET 50-255 - LICENSE DPR PALISADES PLANT - LICENSEE EVENT REPORT 80-021 -

CORRECTIONS TO AUGUST 20, 1980, LETTER By letter dated August 20, 1980, Consumers Power Company transmitted Licensee Event Report 80-021, Revision 1 along with several Appendices.

Appendix C, page 7 did not copy.

Appendix E, page 1 had an error as did Appendix F, pages 1 and 2.

Corrections are indicated by a vertical line in the right hand ~argin. All three (3) Appendices are attached to this letter in their entirety.

David P Nuclear Licensing Administrator CC Director, Office of Nuclear Reactor Regulation Director, Office of Inspection and Enforcement NRC Resident Inspector - Palisades Plant NOV 2 61980

APPENDIX C CALCULATIONS OF RADIOLOGICAL EFFECTS AS RESULT OF HYPOTHETICAL ACCIDENT WITH CV-3030'0PEN

Introduction

Calculations have been performed in accordance with data derived from the analysis given in Appendix D; 490 gallons of containment sump water is delivered over the period between 20 seconds and 150 seconds following a LOCA.

During this period, the recirculation line contains 100% sump water as opposed to 52% as i~ the Appendix C analysis where both ECCS trains were functional.

A second variation from the Appendix*. C calculations 1

arises during the first 150 second~ when the SIRW tank receives a net inflow of fluid because ECCS train A is not functioning to remove water.

The net inflow allows airborne radioiodine release during this period as well as during the two hours after the tank empties to its shut-off level of two feet.

Results Site boundary thyroid dose for the DEA is calculated to be 0.57 mrem due to release of 2.03 mCi dose equivalent I-131.

Site boundary thyroid dose for an MHA is 38.8rem, and results from release of 137.4 Ci of dose equivalent I-131.

Both the DBA and MHA doses are a factor of 1.09 times.the result of the earlier two train analysis. The increase of 9% is due to the release of activity over the 130 second fill period at a release rate pf 1% per two hours (0.000139% per second), given an average concentration of 0.25 curies/gallon over that period.

Dose at the control room console is calculated to be o.44rem.

This increase over the two-train dose of 0.23rem arises from the increase in recirculation pipe consentrations from 52% to 100% of sump concentration in the single train event.

Conclusion Doses continue to represent small fractions of 10CFR50 and 10CFRlOO limits.

A large degree of conservatism remains in the methods by which.these doses are calculated. For fUrther disscusion see the conclusion section of Append.ix C.

2 Results -

MHA Offsite thyroid dose of 35.6 rem at the site boundary is calculated with 25% of core iodine inventory diluted by l.2x105 gallons of fluid available to the sump within the first few seconds of the DBA.

The quantity of iodine released is 126 Curies.

All assumptions are similar to those described for the DEA case.

Dose at the control room console is calculated to be 0.23 rem due to liquids from the sump present in the recircuJ.ation line outside the control room.

Expcsure from the SIRW itself is negligible since d.ilution in SIRW vater is laTge, only 10% of the activity remains after 20 minutes. and the control room is shielded by a minimum of 4 feet of concrete in that di-rection.

Conclusion r;e;i ther control room ha bi tabili ty nor offsi te doses are seriously affected by opening of CV-3030, since in all cases represent small fractions of 10CFR50 and 10CFRlOO limits.

It must be emphasized thc,t doses have been calculated in a conservative manner.

In particular, the MlIA fission product inventory based on TID 14844 greatly excee~s the inventory actually expected in the first few minutes of an accident.

}<'or exaw.ple, WASH-14CO wo:cst-case accident descr:'.. ptions indicate that final gap activity begins to escape the core only after one minute, and core melt occurs only after 16 minutes.*

l)!j 1.J::*

TABLE I (RADIOLOGICAL)

FLUID VOLUMES DILUTING FISSION PRODUCT INVENTORY PRIMARY COOLANT VOLUME 7,800 FT3 1/2 CLEAN HASTE RECEIVER TANK 4,065 FT3 SAFETY INJECTION BOTTLES (li) 4,000 FT3 PRE-EXISTING SUMP VOLUME 304 FT3 TOTAL DILUTION 16,169 FT3 MHA IODINE - 131 CONCENTRATION CORE INVENTORY AT 2650 MwT = 6.65 x 107 CI 5.835 X 104 GAL 3', 041 X 104 GAL 2.992 X 104 GAL 2.275 X 103 GAL 1.210 x ios GAL r

25% CoRE INTO CooLANT Ar T = 0 ~ 1.66 x 107 C1 1-131 -:t> 137 Cr/GAL DosE EQUIVALENT 1-131 ~ 258 CI/GAL I

(

3 e

FIRST 20 SECONDS - DosE EQUIVALENT I-131 EVALUATION (A

---

:---:* ------u"'. q

~--

\\

- ________ _.,;_ ____ ~ __ _:__~~----------------,,

~ I l

~l!l:- 22' 'NATER LEVEL Fi.u;:;;.~cTIVITY Our = AssuMED ZERO

-un. D i NV EN TORY = (' 0.1 CI

--*a>

RELEASE TO ATMOSPHERE*- NEGLIGIBLE

<AIR FLOWING INTO TANK)

(

2 I

I 20 SECONDS TO 150 SECONDS DOSE EQUIVALENT I-131 EVALUATION

' l I I

• I J

0 0

.t' 0

0

...r:'."'

~ 2 2i.

• ~-

Qi

~,__

'- --r..

J r

• a

• rti 1

."?*~... ' * *. : *. *.. * * : *.. * * -

• ,. ~***~-~sefsi@<:f ~-- 2o' vv ATER LEVEL FLUID.~cnvm IN = l.26xl05 Ci ~ 0.50 Ci/GAL 11 ~-~=

~~u1n AcrrvITY bur = AssuMED ZERO

~---*

1**2r 1 5 5

""'~

--~--

h.UID InvENTORY =

,' o;~ 0 CI--> 0, 0 C1/GAL RELEASE TO ATMOSPHERE ~ NEG~IGIBLE (AIR FLOWING INTO TANK)

f\\CTIVITY 150 SECONDS To 20 MINUTES*~ DosE EQUIVALENT l-i3J. EvAu.iArr*oN

---

~,-----------------

0 I

e *

.ii * *

~20'

'I

'I>

'I,

\\

\\

,, 0 1

~

~.

3

~

~... '.** * ** ::--*.. :-...... --:--~

.. ~

i

--i- _. -

-.. ~-

.' ~'<-2' WATER LE.'1'=,

.:*.~-:-

...;n. -,,-......

flO

"_.,-----~

u

\\! t._~

IN ~ ZERO

.~CTIVITY Our = l.13xJ05 C1

= l.26xl04 CI

@ 0.5 CI/GAL

@ 0.5 CI/GAL lS--

INVENTORY RELEASE UiIR TO ATMOSPHERE = ZERO FLOWING INTO TANK)

r.: LU ID 4

20 MINUTES TO 2 HOURS PLUS 20 MINUTES DosE EQUIVALENT I-131 EVALUATION

~

6...

t I J-\\CTIVITY AcTIVITY Our = ZERO INVENTORY = l.25xlo4 CI

.,, *. ~

I I *

---

n I

lb 0

\\

• e ~

'1 f~

e

\\..,

\\

• I I 0 VI/ ATER LEVEL RELEASE JO ATMOSPHERE = 126 CI D,Eo I-131

I I

I I

I I ;

5 DOSE TO THYROID AT SITE BOUNDARY - MHA DosE = {Q) CX/Q) CB) Crl CDCF)

WHERE:

DosE = REM Q

= CI/SEC RELEASE RATE (126 Cz/7200 SEC) = 0.0175 CI/SEC DOSE EQUIVALENT !-131 X/Q

= SEC/M3 DIFFUSION (5,5xlo-4 SEC/M3 PER R.G.1.4 AND AMENDMENT 31)

B

= M3/HR BREATHING ~ATE (l,25 M3/HR PER RsG.1,4)

T

~ HHS OF BREATHING TIME (2 HRS)

DCF

= DosE CONVERSION FACTOR (1.48xl06 RAD/C! INHALED, PER R.G.1.4)

RESULT:

35.6 REM e*

DEA RESULT - PRIMARY COOLANT AT 110JJCI/ML DosE EQUIVALENT I-131: 0.52 MILLIREM

CmnRoL RocM HABITJ\\BJ"LITY G 6" RECIRCULATION LINE EXPOSURE DETERMlNED TO BE ONLY.SIGNIFICANT CONCERN

© 2' CONCHETi: WALL BETWEEN LINE AND CONTROL ROOM 6

• INTERVENING EQUIPMENT AND DISPLAY PANELS NOT INCLUDED AS. SHIEL.PIN~

e 8 15' TO 20 1 DISTANCE BETWEEN SOURCE AND OPERATOR AT MAIN CONSOLE

~ LINE FILLED WITH

1)

MIXTURE 52% SUMP FLUID; 48% SIRW TANK WATER

2)

IODINE CONCENTRATION AS USED WITH.OFFSITE DOSE CALCULATIONS

_3)

PARTICULATE CONCENTRATION BASED ON 1% OF CORE INVENTORY TO SUMP e EXPOSURE DURATION 130 SECONDS RESULT:

0.225 REM DBA RESULT -

<0.01 MREM e*

• ~*.

CONCLUSIONS OF RADIOLOGICAL EVALUATION

• MAXIMUM HYPOTHETICAL ACCIDENT:

1)

SITE BOUNDARY TOTAL BODY DOSE NEGLIGIBLE DUE TO DEPRESSURIZATION OF PRIMARY SYSTEM PRIOR TO REM9VAL FROM SUMP

2)

SITE BOUNDARY THYROID DOSE 36 REM

3)

CONTROL ROOM DOSE 0,23 REM

• EFFECTS ARE CONSIDERED TO REPRESENT SMALL FRACTIONS OF 10CFR50 AND 10CFR100 CRITERIA a DESIGN BASIS ACCIDENT:

1)

SITE BOUNDARY TOTAL BODY DOSE NEGLIGIBLE

2)

SITE BOUNDARY THYROID DOSE 0.5 MILLIREM

3)

CONTROL ROOM DOSE < 0.01 MILLIREM 7

I I

APPENDIX E OFFSITE DOSE AND CONTROL ROOM HABITABILITY ASSUMING LOSS OF TRAIN A SAFEGUARDS PUMPS

~-*

1

Introduction

Since the opening of CV-3030 would not in itself prohibit ECCS operation, it is assumed that only primary coolant activity at the Technical Specification limit of 1 microcurie per gram dose-equivalent I-131 is involved in recir-culation to the SIRW following a postulated DBA.

Two aspects of this event are considered: 1) Release of iodines to the environment via the SIRW tank vent following addition of 490 gallons of sump water to the.tank via the recirculation line; and 2) Dose to control room personnel from the 611 recir-culation line during the 130-second period it is filled with a mixture of 48% SIRW and 52% containment sump water.

For the sake of completeness, doses from MHA fluids circulation also has been determined.

MHA dose parameters are readily available for control room dose calculations, so the MRA calculations were utilized in determination of DBA doses by appropriate scaling of the nuclide inven~ory. Offsite doses were calculated independently by use of iodine inventories for each case. Total body doses from noble gas were not calculated because coolant from the sump vould be lacking in noble gasses due to degassing upon release from the primary system.

Results - DBA

~.aximum offsite dose of.52 millirem to thyroid is calculated from release of 1.86 mCi dose equivalent I-131. This release resulted from a transfer of 490 gallons of undiluted primary coolant to "the SIRW, 100% of which mixes with SIRW vater. The tank empties to a 2-foot level in a maximum of 20 minutes.

No iodine escapes during pump down, since airflow through the SIRW vent is inward at that time.

One percent of the available iodine inventory escapes to the atmosphere within the next two hours (similar to a fuel pool accident described in Regula-tory Guide 1.25). Dose was calculated in accordance with Regulatory Guide 1.25.

Dose at the control room console is calculated to be less than 0.01 millirem integrated over the two-minute period during which primary coolant is flowing through the 6" recirculation pipe outside the control room.

The dose is low primarily because tvo :feet of concrete is present between the piping and control room interior. Dose from the SIRW tank is-negligible because concentration is very low once diluted in the SIRW volume.

Also, a minimum of 4 feet of concrete separates the tank f'rom the control room.

APPENDIX F EFFECT ON MAIN STEAM LINE BREAK AT 1400 MWD/MTU

Appendix F 1

I.

REFERENCES

1.

XN-NF-79-94(p) "Palisades Cycle 4 Startup Predictions and Nuclear Data for Operation".

2.

"Palisades Cycle 4 Startup Data, Supplementary Information",

R.G. Grummer to B.D. Webb, November 30, 1979.

II.

DATA

1.

Net worth (N-1) at 6o°F Reference 2, Table 2 BOC4 = 3.63% ~p,

EOC4 = 4.09% ~p

2.

Power defect at TOO ppmb

- Reference 1, Figure 6.5

= 1-.2% ~p

3.

Net rod worth (N-1) at 532°F

- Reference 1, Table 6.1

4.

BOC4 = 4.90% ~p EOC4 = 5.49% ~p Shutdown boron concentration, keff Reference 1, Table 6.2 BOC4 EOC4 532°F 1000 150

=.98, No xenon -

N-1 Configuration 60°F 1050 500

5.

Core conditions 1400 Mwd./MT Cycle burnup 700 ppm boron

6.

Reciprocal boron worth Reference 1, Figure 6.6 III.

ANALYSIS 1000 ppm, 6o°F, BOC = 77.0 ppm/% ~P 150 ppm, 6o°F, EOC = 70.5 ppm/% ~P

1.

Worth of control rods 6o°F including uncertainty.

BOC

= 3.63 x.90

=

3.27% ~P EOC = 4.09 x.90 = 3.68% ~p Interp~lating to cycle burnup oL1400Mwd/MT worth (N;._l) =

3.27 +

1400 (3.68 - 3.27) = 3.32% ~p 10,400

2.

Reactivity added by cooldown.

This is derived by extracting the change in net rod worth due to cooldown from the change in shutdown boron concentration from hot to cold conditions.

Appendix F 2

a.

At 1000 ppm (1.050 - 1000 - (4.90 - 3.63) x 7fl/77* = -0.62% ~p b.

At 150 ppm (500 - 150 - (5.49

4. 09) x 70. ~ /70. 5

=

3. 56%" ~p c.

Interpolating to 700 ppm

~P700 = - 62 + lOOO - 7oo (3.56 +.62) = 0.86% ~p 1000 - 150

3.

Shutdown Margin - Equals rod worth minus power defect minus reactivity from cooldown.

3.32 - 1.20 - 0.86 = 1.26% ~p CONCLUSION The Palisades reactor could have cooled all the way to 6o°F without boron inejction and remained subcritical until xenon decay.

There is adequate margin in the analysis to account for large uncertainty factors.