ML17194B367
| ML17194B367 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 11/10/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Delgeorge L COMMONWEALTH EDISON CO. |
| References | |
| TASK-15-16, TASK-RR LSO5-82-11-030, LSO5-82-11-30, NUDOCS 8211150167 | |
| Download: ML17194B367 (8) | |
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Docket No. 50-237 LSOS-82-11-030 Mr. L. DelGeorge
- Director of *Nuclear Licensing Commonwealth Edison Company Post Office Box 767 Chicago, Illinois 60690
Dear Mr. DelGeorge:
. -~.
November 10, 1982
SUBJECT:
SEP TOPIC XV-16, FAILURE OF SMALL LINES CARRYING,PRIMARY COOLANT OUTSIDE CONTAINMENT - DRESDEN UNIT 2 Enclosed is a copy of a revised safety evaluation report for Topic XV-16 for Dresden Unit 2. This evaluation has been revised from the previous safety evaluation report dated October 20, 1981, to provide analysis to support the Integrated Assessment and to resolve differences noted* between.the analysis of your facility and two other similar plants.
This evaluation will be a basic input to the integrated safety assessment for your facility unless you identify changes needed to reflect the as-built conditions at your facility. This assessment may be revised in the future ff your facility design is changed or if NRC criteria relating to this subject is,mQdified before the integrated assessment is completed.
Enclosure:
As stated cc w/enclosure:
See next page Sincerely, Dennis M. Crutchfield, Chief Operating Reactors Branch #5 Division ofLLicensing
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NRC FORM 318 (10-80) NRCM 0240 OFFICIAL RECORD COPY USGPO: 1981-335-960
Mr. L. DelGeorge cc Robert G. Fitzgibbons Jr.
Isham, Lincoln & Beale Co~nselors at Law Three First National Plaza Suite 5200 Chicago, Illi~ois 60602 Mr. B.
B~ Ste~henson Plant Superintendent Dresden Nuclear Power Station Rural Rou.te #1 Merri~, Illinois 60450
- The Honorable Tom Corcoran United States House of Representatives Washington, D. c.
20515
- u.. s. Nuclear Regulatory Commission Resident Inspectors Office Dresden Station RR #1 Morris, 11.Tinois 60450
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Mary Jo MurraY Assistant Attorney General Environmental Control Division 188 W. Randolph Street Suite 2315 Chicago, IT1inois 60601 Chainnan
'soard of Supervisors of
.,Grundy County Grundy *county Courthouse Morris, Illinois 60450 John F~ Wolf, Esquire
.3409 Shepherd Street Chevy Chase, Maryland 20015 Dr. Linda W. Little 500 Hennitage Driv'e Raleigh, North Carolina 27612 Judge Forrest J. Remick The Carriage House - Apartment 205 2201 L Street, N. W.
Washington, o. C.
20037 Illinois.Department of Nuclear Safety 1035 Outer Park Drive, 5th Floor Springfield, Illinois 62704 U. S. Environmental Protection Agency Federal Activities Branch Region V Office ATTN:
Regional Radiation Representative 230 South Dearborn Street Chicago, Illinois 60604 James G. Keppler, Regional Administrator Nuclear Regulatory Comniission, Region III 799 Roosevelt Street Glen Ellyn, Illinois* 60137
DRESDEN UNIT ~
SY.f*lATIC EVALUATION.PROGRAM XV-16 RADIOLOGICAL CONSEQUENCES OF FAILURE OF SMALL LINES CARRYING PRIMARY.
COOLANT OUTSIDE CONTAINMENT I.. INTRODUCTION Rupture of lines carrying primary coolant outside containment can allow primary coolant and the radioactivity contained therein toescape to the environment.
SEP Topic XV-16 is intended to review the radiological consequences of such failures.
The review of this topic encompassed those lines \\*1hich carry primary coolant outside containment during po\\.,.er operation.
The scope included those lines that are not normally expected to be open to the primary system but can be opened during power ope rat iop (i.e., reactor coolant sample lines, instrument lines, etc.)
II. REVIEW CRITERION All small lines ca1-ry_ing primary coolant outside containment \\~*ere reviewed
-*-~*-
to ensure that any release of radioactivity from their postulated failure was a small fraction of the 10 CFR Part 100 exposure guidelines.~.*.. Small fraction is defined iri the SRP to be no more than 10% of the lO:CFR Part 100 expos.ure guidelines.
III. RELATED SAFETY TOPICS AND INTERFACES Lines \\*:hich were excluded from this review included lines for \\<Jhich failure outside containment is not postulated, such as lines with isolation val~es inside containment~ or line~ for which i11tcrlocks prevent opening during power operation (e.g. the PWR residual heat removal lines). The review also did not consider the release of radioisotopes from la1-ge pip~s carrying primary system fluid prior to automatic isol~tion of such lines, (e.g. the main steam and feedwater lines of a BWR).
The consequences from failure in these lines are considered in SEP Topic XV-18, "Radiological Consequences of Main Steam Line Failure Outside Containment."
IV~
REVIEW GUIDELINES
- The review was conducted in accordance with SRP 15.6.2.
The licensee 1~as requested to provide plant specific information such as the identification of lines covered by this topic, the size of these lines, break locations and flow, etc.
The licensee responded to this request in a letter dated August 18, 1980.
V.
EVALUATION In th~ submittal, the li~ensee indicat~d that an analysis of the consequences of a I-inch instrument line break was performed ih *
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Amendment 22 to the Final Safety Analysis Report (FSAR) for Unit *3,
- which, the licensee stated, was virtually identical to the Dresden 2 design.
For Dresden 3, the licensee assumed that a break occurs outside the primary containment, but upstream of the check valve, in a 1-inch pipe penetrating the drywel 1.
This break 1~as postulated not to be isolated until after the reactor *was shutdown and depressurized in four hours.
In estimating the amount of primary coolant released, the discharge rate from the break was assumed to decrease during the accident, as action is taken by the control room operator to cool and depressurize the reactor.
This deviates from the SRP which states that the flow is estimated 11with the reactor coolant fluid enthalpy corresponding to the normal reactcir
operating conditions. The time-dependent, decreasing flow rate is justif~ed because the proper response to this accident would be to shutdown the plant to affect repairs.
The staff also assumed that 37% of the discharge fluid flashes to steam and is released tri the environment without credit for Standby Gas Treatment System (SGTS) filtration or plateout in the reactor.building.
In addition, an iodine spike was assumed to occur as a result of the reactor shutdown or depressurization of the primary system. The spike was modelled by increasing the equilibrium iodine release rate from the fuel by a factor of 500.
These assumptions are in accordance with the Standard Review Plan.
The Dresden primary coolant technical specification limit for iodine activiy consists o_f a single shutdown value of 20)JCi/gm gross :iodine aci:i'.vity.
This 1 imit conflicts with the two-tier (equilibrium arid spike). ~ose equivalent (D.E.) I-131 limits found in the BWR standard technical specifications (STS), and also with Standard Review P;lan Section 15.6.2.
However, assuming that the shutdown limit is tornposed of entirely 1-131 and using this single shutdown limit as the ~~uilib~
rium li~it, the calculated doses well exceed the dose guidelines'of 10 CFR Part 100.
The doses listed in Table 2, how~ver, were calculated assuming the STS equilibrium value for primary coolant iodine activity.
Implementation of the STS" significantly reduces the calculated radio log-ical consequences of this accident.
VI.
CONCLUSION The results in Table 2 show that even with the adoption of the BWR STS D.E. I-131 reactor coolant equilibrium activity limit of 0.2J.1Ci/gm, the
EAB dose to the thyroid is 128 Rem which exceeds the SRP guideline value of 30 rem.
Because the calculated thyroid dose is directly proportional to the D.E. I-131 reactor coolant equilibrium activity, the SRP guideline value of 30 rems, thyroid, can be achieved by reducing this activity from 0.2
)JCi/gm to 0.047 pCi/gm, D.E. I-131.
In any case, use of the*STS value results in doses which are less than the guideline values of 10 CFR Part 100.
It should be noted that the evaluation performed by the staff was based on the SRP Section 6.2.3 guidance in Branch Technical Position 6-3 that 11Hhenever the pressure in the secondary containment volume exceeds -0.25 inches w. g. (
water gage), the leakage prevention function of the sec~ndary containment is assumed to be negate*d.
11 Hm<1ever, some credit
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maj be justified for the Standby Gas Treatment System (SGTS} in mitigat-ing the radiological consequences of this accident by consideration of
- 1) t_he integrity~ of the secondary containment under the positive pres-
- surization caused by the line break, and 2) the location of the broken line with respect to both potential leakage paths through the boundary of the secondary contain~ent and the intake(s} to the SGTS.
Such credit could be purs~ed 'through the integrated assessm~nt and has the potential for considerable reduction in the calculated doses.
- e.
TABLE 1 Assumptions Used in the Radiological Consequences of Instrument Line Break Outside Containment at Dresden Unit 2
- 1.
Mass of reactor coolant in vessel-mixing volume (lbm) 590,000
- 2.
RWCS cleanup rate {gpm) 600
- 3.
Condensate demineralizer cleanup rate (carryover fraction x feedwater flow rate) {gpm) 263
- 4.
Iodine spiking factor 500
- 5.
Flash fraction {percent) 37
- 6.
Duration of accident (hours) 4
- 7.
X/Q values Ground level values 0-2 hr, EAB 0-4 hr, LPZ 3
(sec/m )
s~- Reactor coolant concentration 'fCi/gm)
- 9.
bischarge rate of reactor coolant Time Afte~ Break
- : from break hours 0-0.5 0.5-1.0 1.0-1.5 1.5-2.0 2.0-2.5 2.5-3.0 3.0-3..5
. 3.5-4.0 2.6 x 10-4 1.1 ~ 10~5 0.2 Discharge Rate lbr.i/hr 96;000 87,000 69,000 53~000 3T,OOO 25,000 14; 000 9;000
- 10.
No Credit for Standby Gas Treatment System filtration
- 11.
RWCS continues to function during the accident
- 12.
No cleanup from condensate demineralizer following the break
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- e TABLE 2 Radiological Consequences of the Instrument Line Break Outside Containment at Dresden Unit 2 0-2 hour, EAB 0-4 hour, LPZ
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Thyroid Dose (Rem)
Whole Body Dose (Rem) 128 9
0.02 0.002