ML20090B485
| ML20090B485 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 05/12/1982 |
| From: | Houston R Office of Nuclear Reactor Regulation |
| To: | Novak T Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML082380335 | List:
|
| References | |
| FOIA-91-106, TASK-2.F.1, TASK-TM NUDOCS 8205210429 | |
| Download: ML20090B485 (28) | |
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b MAY 121982 Plant File PEasley W9asedag Docket No, 50-244 LGHulman FCongel WGamill RWHouston (AD/RP RF C)
MEMORANDUM FOR: Thomas M. Novak, Assistant Director e
c' for Operating Reactors SECgg.ED Division of Licensing
%y28 FROM:
R. Wayne Houston, Assistant Director for Radiation Protection c,
A.
Division of Systems integration
SUBJECT:
RADIATION PROTECTION INPUT TO GlhNA RESTART SER ro r
Enclosed is the input from the Effluent Treatment Systems Franch (ETSB),
Radiological Assessment Branch (RAB) and Accident Evaluation Branch (AEB) for the subject safety Evaluation Report.
The ETSB evaluated the effluent monitoring system function during the Ginna accident, as descrined in Section 5.6 Based on tneir evaluation, the ETSB concluded that the functional failures of the high range noble gas effluent monitors (the monitors are required for HUREG-0737, Iten 11.F.1., Attachnent
- 1) during the steam generator tube rupture accident do not preclude a re-start-of the Ginna reactor.
However, we do recornend that the licensee develop an operability surveillance program and reevaluate alarm setpoints for these monitors.
J. Lee prepared this evaluation.
The AEB prepared Sections 7.1 to 7.3, and 7.8.
P. Easley was the primary technical author. Staff concerns regarding radiological consequences relative to the design-basis steam generator tube rupture accident are addressed in Section 7.1.
The Cornission concern regarding the acceptability of the auxiliary building ventilation system intake location is addressed in Section 7.8.
The Cocnission question regarding the.,5ecking of secondary coolant activity prior to the intentional release from the unaffected stean generator is addressed in Section 7.2.
The discussion of radiological consequences in Section 7.1 expresses concerns about the ability of the licensee, using present and proposed short-term revisions to the emergency operating procedures, to control the duration of the accident and the level in the affected steam generator.
Therefore, stringent Technical Specifications are required on the reactor coolant activity limits and surveillance requirements. The factor of five reduction (relative to the Westinghouse Standard Technical Specifications) in the limits on dose-equivalent I-131 would cover these potential non-l conservatisms until the licensee provides an analysis, acceptable to the,
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f Thomas M. Novak staff, that justifies higher lintits, not to exceed the Westinghouse Stan.
dard Technical Specification limits. The analysis should be performed within six months of the issuance of the safety Evaluation Report. The analysis should, among other things, address procedure changes to prevent prolonged primary-to-secondary leakage and steam generator overfill. The Accident Evaluation Branch has coordinated with the Reactor Systems Branch on the requirements for this analysis.
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The RAB evaluated the licensee's environmental monitoring actions during the course of the Stean generator tube rupture accident, as described in the attached 7.4 to 7.7.
Their evaluation found that the licensee's
- tions were, in general, consistent with good health physics practices and that the licensee's interpretation of data and conclusions are con-sistent with ours. However, the R AB recomends that the licensee develop a specific procedure for the uniform collection of snow samples during other than normal atmospheric releases of radioactive materials for inclu-sion in the final energency response plan. Restart of the reactor should not be delayed for the approval of this procedure. Further, the R AB con-cluded that none of the licensee's actions in the area of environmental monitoring during ths. event preclude a restart of the Ginna reactor.
H. Wangler and J. fienenias prepared this evaluation.
Criginn '.tr.e4 by' R. hyne husten R. Wayne Houston, Assistant Director for Radiation Protection Division of Systems Integration
Enclosure:
As stated cc: H. Denton W. Pasedag D. Eisenhut T. Quay R. Mattson M. Wangler P. Easley J. l.ee B. Sheron J. Nehemias
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v GINNA RESTART SAFETY EVALVATICtt REPORT Section 5.6 and Chapter 7 A
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5.6 3ffluent Radioactivity Monitorino system The release pathways for sirborne radioactive t-a t e r i a l s from R. E.
Ginna Nuclear Power Plant to the en.ironment during the steam generator tube feiture incident involved three af fluent radioactivity mo ni to ri ng systems (1)
The main steam radi ation monitoring syetem, which is designed to de te et, indicate, record, alarm, and quantify radioactive materlats released from the steem ge ne ref o r PO RVs and safety valves; (2)
The air ejector radiation mon 4 to ri ng system, which is designed to detrat, indicate, record, alarm and queurify retrases of radioactive materials in nonconde ns ib le
. gases frts the secondary Jystem steam via the air ejecter end turoine gland seat e xh au s t; and (3)
The plant ve nt i'.a tio n a f f lue nt mo ni to ri ng system, which is designed t o moni to r radioactive particulates, noble g a s e s',
and radioiodines in ventila tion air di sch arge from the Auxili ary Buildi ng.
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I 5.6.1 Main steam Radiation Monitorina System The-system consists of a collinated energy-compenrated Geiger-Muette r detector (Ebe rline Mode L S A-11) on each main steam li ne.- The system indicates r adi o a c t ivi ty readout localty and in the main control room.
A high r adi o act ivi ty alarm activates the system recorders to start continuous recording of radioactivity i n the main. steam line and the s team ge ne rato r PORV and safety valve positions.
Radioactivisy releases can then be quantified by taking the product of steam flow rate, radioactivity concentration, and the time duration that the valves were open.
-This system was _ installed in Decembe r 1981 to satisfy the requi rement s.in NUREG-0737, It em II F.1, at tachment 1, high range noble gas effluent monitor.
During the incident, however, a high radi ation ala rm setting was
- not reached-.and this prevented the system fron &ctivating the I
recorders.
Later attempts to retrleve the data from the
. moni toring data.proces sing sys tem also f ailed dire to a malfunction of'the monitor during the incident.
The licensee states in his i n ci de nt evaluation report that the monitor mutf unction is.
4 believed to have been due to a small smudge of dirt or residue which caused electrical teakage on a printed ci rcui t board.
In addition, the steam gene rator PORV and safety-valve position mo ni to ri ng function also failed during the i n ci de nt.
The licensee a
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3 states that inadequate adjustment of the new actuator rods installed on the safety valves, and open sliding links on terminal blocks in the elay room for the PORV, caused the inoperability of the valve position monitoring.
Within three nths subsequent to the plant restart, the staff
.4 will complete the review of (1) the adequccy and basis of the monitor high alarm setpoints, (2) the monitor opersbility surveillance program, (3) the monitor ranges and sensitivity with respect to their capability to cover the e'n t i r e range of effluents from normal (ALARA) through accioent conditions, and and (4) procedures or calculative methods to be used for converting monitor readings to release rate per unit time.
5.6.7 Air Ejector Exhaust Monitoring System
-The system consists of two radiation monitors: the R-15 monitor, and the SPING R-15A monitor.
The R-15 monitor is a sodium iodide detector (Victoreen Model.No. 843-03) mounted on the outside of tne 8 inch diameter exhaust pipe and has been in service since 2
6 1979
'he monitor has a range of 10 to 10 cpm gamma radiation (equivalent to O to 0.1 pci/cc).
The response of this monitor is recorded on a strip chart ar4d also fed to a computer.
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4 During the incident, the strip chart recorde r f or the R-15 monitor went off scale f or 105 seconds beginning at 0926 of J anua ry 25, 1982.
The SPING R-15A monitor (Ebe rline Model SPING-4) is a high range moni to r and has three sensitivity ranges with a separate detector for each range.
Range Deteds or Range, pCi/cc
-6 tow beta sci nt illa tion 10 to 0.05 middle compensated GM tubo 2.8 x 10 ' t o 10
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10' high compensated GM tube 0.03 x Ouring normal operation, only hourly ave rages of the monitor readout s are p rinted out with the capability of p r ov i d i ng an inrtantaneous readout on demand.
Only a high radi ation ala rm activates the system to provide 10 minute average readouts and recordings.
This monitor was ins t al led i n D e c e mb e r 19 81 to satisfy the requirements in NUREG-0737, Item I I. F.1, A t t a c h m e nt 1, neble gas effluent monitor.
During the incident, the SPING R-15A low range monitor actuated a high radiation alarm and is suspected to have been of f scale,
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5 after activating the SPING R-15A middle range monitor.
- However, while the low range monitor was off scale, no 10 minut e ave rage radiation readout s o r recordings were obtained from the SPING R-15A middle range monitor because a high alarm setpoint was not reached and this prevented the system from activating the recorder.
Subs eque at to the incident, the licensee made corrections on these alare point settings such that an alarm on the. low range monito r actus tes t ha 10 m i nu t e average reaceuts w
on the middle range monitor.
The licensee should provide a continuous ar.d instantaneous i ndi c a to r-r ec orde r (strip chart) in addi tion to t he 10 mi nut e
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average rendouts in the computer to indicate release rate of airborne radioactive n,sterials from the air ejector exhaust to the-environment.
Within three months subsequent to the plant restart, the scaff will complete the review of (1) the adequacy of readouts and recording capability, (2) tre dequacy of all monitor alarm setpoints, (3) t he monito r ope rabili ty surveilla nce progr am, and (4) the procedures or calcula tive methods to be used for i
converting the monitor readouts to release rate per unit time.
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6 5.6.3 Plant.. Ventilation Exhaust Monitoring System The monitoring system consists of radioactivt partir.utate, noble gas, and radioiodine detectors.
A continuous sample is drawn from the main exhaust vent stack.
The system has been in service since the plant startup.
These monitors functioned p rop e r ly throughout the tube rupture incidert and provided important information regarding the timing and amounts of radioactivity releases due to the safety valve liftings.
The intake for supply. air to the Auxiliary Building is located on the roof of the Auxiliary Building at a point which was, most of the time during the incident, downwind from the safety valve
- releases.
Therefore, the plant vent monitors detected the radioactivity released from the safety valve drawn into the Auxiliary Building through the supply air intake.
The monitor readouts reflected the time and duration of the s a f e t.w valve liftings and releases.
The adequacy concerning the location of supply air intake to the Auxiliary Building is discussed ir Section 7.8.
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7.0 Radiological Assessment During the J anuary 25, 1982 steam generator tube rupture accident ($GTR) at Ginna, radioactive primary coolant leaked to the 8 steam generetor.
Some of the contaminated secondary coolant was than released to tha environment.
A description of the releases and the follow-up activities is presented i n i
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chapter 5 of NUtt,-0909 and Chapter ? of the iscinsee's
" Incident Evaluation."
A Jiscussion of the reseeses and the sonitoring, surveying, and sampling activities after the accident, as they relate to Ginna restart, cad recommendations for licensee actions Srior to and after restart, are presented in the following sections.
I 7.1 Recommendations for M131ostion,_of Radiological *onseque,cet During the January ?.5, 1982 accident at Ginna, the total amount of primary ~to-secondary leakage and the total amount of water and steam released to the environment,ere larger than vould normally be predicted, because of valve malfunctions and operator actions (see Chapters 3, 4, and 5 of NUREG-0909).
A comparison vith a previous safety evaluation report input on th* radiological consequences of a steam generatnr tube rupture ccident ($GTR) (f or the Syst ematic Evaluation Program, W.
E.
eger memorandum to G. C. Lainas, June 25, 1981) shows that the potential exists for doses exceeding Part 100 Gui,1elines from a design-basis SGTR accident.
There doses would occur only i f there were an unlikety, but not impossible, set of circumstances, namely:
primary coolant with todine concentration at tha Westinghouse Standard Technical 9pecification coolant iodine concentration spiking limit of 60 pCi/1 dose-equivalent I-131, maximum flow rate through a double-ended tube rupture, flow through the tube rupture prolonged for two or more hours, i
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8 filling of the steam generator and steam line of the aV1ected steam generator, releases through the effected ste6m generator's safety o f-atmospheric dump / relied valvee a s a tvo phase mixture, and conservative atmospheric dispersior, factors.
The actual radiological consequences of the Ginna accident were not severe because the reactor coolant iodine concentration wos very low, 0.057 Ci/g dose-equivalent 1-131; and because the meteorologic conditions were far more favorable, with respect to offsite doses, than the conservative assumptions used in the prior analyses.
Some aspects o' the Ginna accident were as severe as some of the a b o v.' aseutptions:
the high initial rupture leak rote, the prolonged leak, the filling of the steam generator arid part of the steam line, and the release of a tvo phase mixture through the safety valve.
Although a more verious ac'ident vas avoided and the radioactivity releases were not excessive, the staff believes that corrective measures must be iaken to prevent potential accidents in the future from having similarly large leakages and releases that could cause more severe radiological consequences.
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9 The f ollowing discussion is not meant to describe the ac cident (as do NUREG-0009 and the licensee's
"!nci de nt Evaluation"), but to evaluate the generat implications of events at they might affect the radiological consequences of potential accidents, and to provide a background for the suggested corrective measures.
T vo t ',ings t hat contributed to t he p rolonged primary-to seconda ry le ak a ge and the ove rfit ting of the steam generator we re (Indirectly) the pressurizer PORV that stuck op e n, a nd t h e c!c lay ed decision to terminate s a t. l. y injection (the Jtaff notes that there were ot he r legitimate concerns that required evaluation, that may have contributed to this delay).
Ove rfilling of the steam generator 12 undesirable with respect to releases of radioactivity for several reasons: as the water level rises above its normat value, the steam dryers flood, whi:.
permits higher moisture carryover (li q ui d d r op le t s that may amount to a few percent of the total mass re le as ed) with the steam.
When the steam ge ne r g to r is ove rfilled ano t he steam tine is flooded to the M51V, i* is likely that most or all of the rel e ases from the reilef or safety valves will be as water or a two-phase misture.
This can cause a mais
l 10 flow rate of contaminated secondary coolant to the environment that is highsr than the design A
release rate, which is based on steem.
(Two-phase flow thro ei *te relief or safety valves may contribute to valve orgradatien and possible failures to reheat.
This can contribute to ths: radiological consequences by providing a prolonged pathway to tha environment.
The evaluation of both the pressurizer PORV and the safety valve function and their repair, is elsewhere in this SER.)
The release of a two phase mixto e also results in the transport to the environment of non-volatile radionuclides not expected to be present in a steam-only release, and a much higher concentra+ ion of iodine in the material released, because of the lack of liquid phase / gas phase partitioning.
The release of radioactive steam snd water to the environment to relieve the pressure in the affected steam generator was worsened by the decision to close the block valve upstream of the affected (B) steam generator atmospheric dump / relief valve, and the failure of the B steaN generator safety vatve to reseat fully.
The block valve closure forced the safety valve to Open and close repectedly to relieve pressure over a period of about two hours.
This cycling, o
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11 or the releese of two-phase wster/steem, say have contributed to the failure of the safety valve to ressat fully (s ee pages 3-18 an a 3 19 o f NUREG-0909 f or f urther discussion).
During a period of prolonged leakage into the steam genorator, for which pressure relief may be repeatedty required, it day be better to use tne steam generator atmospheric dump / relief vatve on the affected
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steam generator, since that type of valve is better suited for eycling than safety valves.
The licensee han recommended (s ee S e c t ion 8.1 o f the Incident Evatuation") several short-term procedural changes.
A review of all these recormanded changes appears elsewhere in this SER.
The following specific changes or additions to Procedures L 1.4, "Jteam Generator Tube Rupture," as proposed and annotated by the licensee, are designed to increase the likelihood that that unnecessary demand on the steam generatc* safety valves, prolonged primary-to-secondary leakage, and steam generator overfill vill not 0: cur
" STEP 3.9.3 thange te read, 'put atmospheric steam dump controller in the manual closed position'.
This change will clarify that the controller is to be put into manual, but that the CblockJ valve itself need not be manually closed.
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" Ado NOTE
' Termination of SI with suspected After STEP voids in the upper RV head is allowed 3.15.3 when natural circulation is verified.
(Refer to 0-8)'
This note serves to eliminate hesitancy in terminating 51 when the 51 termiration criteria are met and natural circulation 1$ assured.
" Add STEP
' Block $1 before the f aulted S/G drops 3.20.3 below 550 psig '
This will prevent 51 re initiation due to lou SG pressure '"
The licensee is also recomending a long-term procedure change:
" Add a section to the procedure to address operationwiththe[aultedsteamgeneratorfullofwater."
However, the recomended procedure changes are, by themselves, insufficient evidence to the staff that potential steam generator tube rupture accidents will not result in offsite doses excee} ting Part 100 guidelines.
In particular, all of the recent SGTR accidents have shown that with the reactor coolant pumps (RCP) tripped, it has taken longer to equalize primary and secondary (affected stem generator) pressure than the licensee assumed in their FSAR. The licensee for Ginna has not proposed any revision to their present RCP trip criteria which would allow the RCPs to remain operationt,1 if tube rupture similar to the January 25 rupture recurred.
Moreover, with the RCP5 tripped, the potential for primary system void formation, overfilling the steam generator, and two phase discharge from either 3 safety or relief valve is increased.
Thus, until the licensee cas develop an RCP trip criteria or install modifications that will prevent RCP trip for SGTR accidents for which the trip is not required for safety purpose, corrective measures are prudent and warranted. Because the licensee's analysis did not consider some of the facturs noted above, and because of the concern about potentially high doses and the incomplete evaluation of the effects of changes
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13 to operator pJidslines, the staf f recommends that the licenses re-analyse the radiological consequences of an SGTR accident.
The new analysis should, in particular, either analyse the ef f ect of prolonged primary-to-seconday leakage and overfilling of the steam generator, or provide evidence that this will not occur. The staf f is also reviewing, generically, the Standard Review Plan for the radiological consequences of an SGTR. As an interim measure, to redJce the probability that a steam generator tute rupture in the future will not result in severe radiological consequences, the staf f recommends that the technical specifications f or primary coolant iodine concentration be changed, prior to restart, as f ollows: 1) a limit on the maximum primary coolant activity dJring spiking of 12 pCi/g dose-equivalent 1-131, (DE 1-131), a limit which, if exceeded, requires the plant tu be in Hot $tandby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,
- 2) a long-term, equilibrium primary coolant dodine concent ration of 0.2 pCi/g DE 1-131, which cannot be exceedec more than 48 consecutive hours without placing the reactor in Hot Standby, 3) a limit on the total duration during which the equilibrium limit is exceeded of 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> per year, and 4) the sampling and reporting requirements specified in the Westinghouse Standard Technical Specifications for those times that the coolant iodine coricentration e sceeds or has a potential f or exceeding the equilibrium limit. The staff proposes that, if there is a f avorable review of the recommended licensee SGTR evaluation, these technical specification limit may be replaced by the less stringent Westinhouse Standardard Technical
$pecification (STS) li mi t s L__
(~'t 14 for iodins activity, retaining the surveillance requirements.
The licensee has agreed to change some of their technical specifications to conform to the Westinghouse STS following the Systematic Evaluation Program Integrated Assessment (J. E. Maier letter to 0.
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Crutchfield November 4, 1981).
The staff notes that it took the operators '5 minutes to Identify positively the steam generator with a tube rupture during the Ginna accident.
With respect to radiological consequences, the staff concludes that this was not an eacessive time.
For design basis analyses, the staff typically assumes that 30 minutes are required for positive identification of the affected generator.
74 Releases from the Unaffected Steam Generator i
It has been noted that the dumping of slightly contaminated stead from the unaffected (A) steam generator amounted to an intentional release to the atmosphere, this is a necessary and normal response to a steam generator tube rupture when the condenser is not available; and, because s
there are a variety of reasons why it is impossible or undesirable to use the condenser following the accident, this is the case for which licensing accident evaluations are 1
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However, the condenser is likely to be avaitable, and useful, after most accidents.
It appears that, during the January 25, 1902 accident at Ginna, removing the condensers from service could have been more carefulty evaluated, taking into account the alternatives and the effects of additional i
environmental releases (see p. 3-15 of NUREG-0909).
Recomasndations have been made elsewhere in this Report for procedure changes to use the condenser as a heat sink (in conjunction with the unaffected steam generator) to effect cooldown folLowing an SGTR, in preference tv the atmospheric dump valve.
During the Ginna accidents the Licensee did not sample the A steam generator water prior to the first releast from the A atmospheric dump valve.
7.'3 Meteorology The staff has no objections to, or conditions on, the Ginna restart with respect to meteorology eonsiderations.
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In Section 7.4 of the licensee's incident Evaluatinn Report and ir l
l section 5.7.5 of NUREG-0909 the actions and findings of the licen-tee's survey teams are discussed.
During the event the licensee dispatched 2 onsite and 3 offsite 2
survey teams to record direct radiation caposure data and to collect env'ronmental sample
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aire water, and snou).
Each team was equipped with Geiger-Mueller and scintillation detection equipment.
Each team was assigned a particular route and conducted at least l
2 surveys of-each-roete during and subsequent to the relenses.
The surveys were conducted primarily in the quadrant SE of the release point to a distance of 4 miles.
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The highest exposure este measurement made in t he course of the surileys, 3 mr/hr, was obtained at the site fence, approximately 130 meters southeast of the release point, as the team passed under a radioactive steam cloud.
All other measurements onsite were lower by a factor of 2 or greater.
Beyond the plant boundary all radiation levels were at backgroand levels with me exception v
of one measurement of 1.2 mr/hr near the plant entrance.
2 These data were used by the NRC staff and the licensee to evaluate possible exposure to the maximum-exposed individual offsite.
The NRC staff considers licensee actions in this area to be consistent with good health physics practice.
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7.$.1 Air Sampling The licensee's fixed air sampling stations were operating through-out,the event.
Survey teams dispatched by the licensee shortly after the beginning o.' the event also collected airborne contaml-nation samples with portable air sampling equipment.
Three sampling locations exhibited iodine concentrations above background.
These samples were collected downwind from the plant.
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On the day of the event, instantaneous iodine concentrations higher than the annual avtrage permitted by 10 CFR Part 20 for unrestricted areac were measured at one onsite plant location.
However, within 4 days of the event iodine concentrations had decreas,td to less than the detection limit of the counting equipment, i
At 2 offsite locations the measured radiciodine concentrations were less than 25% of the maximum permisslble air concentr.tions for unrestricted areas as specified in 10 CFR Part 20.
Because the iodine con:entrations remained above background for several days after the releases, the licen see concluded that volatile radioiodiaes deposited on buildings near the release l
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The NRC staf f considers
- that the licensee's conclusions are consistent with the data, and that licensee actions in tnis area during the event were consiste + with good health physics prac-tices.
LS.2 Snow Sampling _
During the event snow was falling at a rate of about 1/4 inc.h per hour in the vicia.ity of Ginna.
The licensee collected more than 100 snow samples from the ground, from vehicles, and from buildings.
Recause of the wide variability in area, depth, density of the snow samples collected, ano porsible cross contamination of s a rn -
ples, measured concentrations of radioactive materials in snow could not be used to determine deposition quantitatively.
- However, a comparison oi the relative concentrations for snow collected onsite and offsite indicates that a major portion of radioactive materials was deposited within the site boundary.
The samples further indicated that a significant proportion of the iodine available for release f r om th e a steam generator was deposited
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The-NRC staff considers the licenseds conclusion about the rela t iv e depositon of radioactive materials to be consistent with the data.
However, the NRC staff has concluded that.if the li c en s ee survey teams had established and implemented a uniform snow sampling
-procedure, an accurate assessment of the deposition of radioactive material could have been made.
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f 19 Therefore,the staff recommends that the licensee develop a spe-cific procedure for the unifora collection of snow samples, under snow conditions, during other than normal atmospheric releases of radioactive materials.
7_.S.3 Water S a mely Onsite tap water samples and offsite samples at the Ontario Water Works were taken by the licensee.
None of the analysed indicated radionuclide co'ncentrations above the minimum detection capability of the instruments used.
The NRC staff _ considers licensee actions in this area to be consis-tent with good health physics practices.
7.6 TLD Meesurements The licensee had placed thernolumimescent dosimeters (TLDs) at 32 offs ~ite locations including 11 at the site boundary and at 7 on-site locations.
Nine additional TLD's were placed offsite by survey teams immediately following the event.
Additionally, the NRC had 27 TLD's offsite and the State of New York had 2 TLD's ensite.
With the exception of the two TLD's ensite that were situated approximately 0.2 mile downwind SE of the release point, no TLD measurement indicated an exposure significantly above background.
The 2 TLD's that recorded significant exposures (21.7 millirems as measured by RG5E and 9.4 millirems as measured by New York State) were located at the approximate centerline of the predicted g
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Most of the these TLD exposures probaoly came from radioactive materials deposited on the ground and nearby surfaces esther than from the plume itself.
These data'were used by the NRC staff and the licensee to evaluate pocsible exposures to the maximum-exposed individual onsite.
The NRC staff considers licensee actions in thix area to be consistent with good health physics practices.
7_7hstimatedoffsite Doses l
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In Section 7.7 of the licensee's incident Evaluation Report and in section 5.8 of NUREG-0909 the offsite population and maximum-
'T exposed individual doses are discussed.
External exposure, inhala-tion, and ingestion pathways were considered.
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for plume exposure the licensee investigated the macimum individual doses from 2 sources of radiation, inhalation of radionuclides and t
external exposure to radiation.- The licensee concluded that the maximum-exposed individual could have re'.aived a thyroid dose of 2 millirees offaite and a whole body dose of 0.07 millirem.
The
-NRC staff has estimated that the msmimum-exposed individual could have received a thyroid dose of less than 5 millirems of f site and
'a whole. body dose of 0,5 millirem.
The difference in the whole body rioses may be at t rbut ed - t o t he NRC staff's higher estimated source terms (cf Table 5.4 of NURCG-0909 and Tables 7.2-4 and 7.2-5 of the licensee's report).
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n 21 For population dosos the licensee estimated that the maximum whole is 0.2 person-body population dese withic 40 miles of the plant The NRC staf f has estimated that for the population within rem.
a 50 mile radius of the plant, the whole body dose is less than 0.1 person-rem.
Additionally, the licensee considered potential i ngestion pathways, such as fish and drinking water consumption, due to runoff of melted snow into Deer River and the lake.
The licensee concluded that the maximum-exposed individual would receive about 0.6 milli-rem and the maximum population dose wouto be 1.3 millirems.
How-
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ever, the licensee feels that these doses overestimate the actual doses because-a large fraction of the depo 2ited radioactivity in c
snow was plowed w and stored in the Rodwaste bunker to allow radioactive decay prior to entering unrestricted water bodies.
The staff has made no comparable analysis because of the lcck of radionuclide deposition data.
Finally, the licensee compared predicted dose rates based on esti-i mated radionuclide releases with actual measured dose rates.near plant buildings.
The licensee has determined that the actual dose rates were lower than the predicted dose rates.
The NRC did not perform o comparable analysis but considers the licensee's
-findings to be consistent with the data.
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n Based on the foregoing discussion the NRC staff has concluded that the licensee evaluation of offsite doses are consistent with that of the NRC staff.
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23 7.8 Recommendations Regarding Ventilation Intake During Steam Generator Tube Rup,ture Retsalsg During the January 25, 1982 accident at Ginna, air contaminated by stems and/or water droplets released from the affected stese generator safety valve was pulled into the auxiliary building through the ventilation intake.
(This is discussed further in pages 7.2-4 and 7.2-5 of the licensee's " Incident Evaluation".)
Although relocation of the intake may be unrealistic, the staff recommends that the licensee consider a procedural change calling for closure of the auxiliary building's (and perhaps other buildings') ventilation intake ports, or turning off some of the intake fans, while a steam generator with a tube i
rupture has open safety or relief valves.
The evaluation
. of-this change should consider potential doses fron the safet,y/ relief valve source, potential doses resulting from disturbing normal ver,tilation flow paths for a short time, and potential short-term reductions in the cooling of safety-grade or safety-related equipment rooms.
(The staff notes that the Ginna FDSAR states that essential equipment in the auxiliary building is supported by separate cooling and ventilation sy1tems.)
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e 24 7.9 Summary of Recommendations i
In the previous sections, the staff has recommended severst changes to sitigate radiological conseq~ races from an SOTR i
l at Ginna.
In summary, they ares to adopt nev technical specifications for reactor coolant iodine activity concentration and surveillance requirements providing for lower limits; to make procedural changes to reduce the chances of unnecessary safety valve use, prolonged primary-to-secondary leakage, and steam generator overfilling; and to consider procedural changen to prevent or lessen ventilation intake of cont ami na t ed a i r du c t'ng accidental releases.
Also, the NRC staff has evaluated the licensee's enviromental monitoring program during the event and finds that the Licensee's actions wera, in general, consistent with good h.entth physics practices and that the licensee's interoretation of data and conclusions are consistent with those of the NRC staff.
However, the staf does recommend that the licensee develop a specific procedure for the uniform collection of snow samples during other than normal atmospheric releases of radioactive materials for inclusion in the finst emergency response plan.
Restart on the reactor shou,Ld not be delayed for the development of this procedure.
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