ML19259D287
| ML19259D287 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 09/18/1979 |
| From: | Donohew J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19259D279 | List: |
| References | |
| NUDOCS 7910170570 | |
| Download: ML19259D287 (12) | |
Text
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UNITED STATES OF AMERICA fluCLEAR PEG"LATORY CO.'041SSIOil DEFORE THE A10:11C SAFETY AND LICri 5ING COARD In the Matter of
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DAIRYLAND POWER COOPERATIVE
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Docket No. 50-409
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(SFP Licens2 Amendnent)
(La Crosse Boiling Water Reactor)
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AFFIDAVIT OF JACK fl. DON 0 HEW CONCERNIflG BOARD QUESTIONS B.1, B.2, B.3.a, b, c; C; D; F.1 L.2; F.3 F
My name is Jack N. Donohew.
I am assigned to the Environnental Evaluation Branch of the Division of Operating Reactors within the United States Nuclear Regulatory Commission. The purpose of this affidavit is to respond to certain Board questions submitted to the tiRC Staff and Applicant by the Board's Memorandum and Order of September 7,1979.
My professional qualifications were previously submitted as part of the flRC Staff Motion for Summary Disposition filed. July 30, 1979.
Questions B.1 and B.2 The Staff's SER indicates that the estimated potential consequences of a fuel handling accident are 162 rem to the thyroid and 2 rem to the whole body at the exclusion area boundary and smaller within the low population zene (LPZ)
(SER,H3.6.1).
This analysis apparently a'ssumes that freshly discharged fuel assemblies will nct be stored in an upper rack position directly over another freshly discharged fuel assembly. The stated consequences are within the guidelines of 10 CFR Part 100.
Even if a freshly discharged assembly is stored-over another such assembly, the offsite consequences are said to be within the expo ure nuiAlin' of 10 CF') P,rt 100,.ithounh n; ex-posures are set forth..Given these circuistances:
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$22 gotorro679
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1.
Will the Applicant be prohibited by a technical specifica-tion or other requirenent from storing a freshly discharced fuel eleme6t over another freshly discharged element?
2.
If not, what are the estimated maximum offsite consequences of an accident involving a freshly discharged element in place di-rectly over another freshly discharged element?
Response
(B'.1)~ The applicant will not be prohibited by a technical specification from storing a freshly discharged fuel assembly over another freshly discharged fuel assembly in the LACBWR double tier spent fuel pool.
We have discussed the potential radiological consequences of fuel handling accidents in the LACBWR pool, including the case of freshly discharged fuel assemblies being stored in an upper rack position di rectly over another freshly discharged fuel assembly, in Section 3.6 of the Safety Evaluation for the proposed pool modification dated July 1979, and have concluded that the potential consequences of such a fuel handling accident would be within the exposure guide-lines of 10 CFR Part 100. The licensee can choose to prehibit this.
storage configuration by means of adainistrative procedures.
(See the response to Question C also.)
i162 jz3
. (B.2) The potential offsite radiological consequences of postulated damage of the described three freshly discharged fuel assemblies in the LACC'.lR pool, during fuel handling are 243 Rem thyroid and 3 Rem whole body at the Exclusion Area Boundary.
The potential radiological consequences of this postulated accident at the Low Population Zone will be sub-stantially lower.
These potential consequences of the low probability event are within the exposure guidelines of 10 CFR Part 100.
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Question B.3 The Staff's SER indicates that the estimated potential consequences of a fuel handling accident are 162 rem to the thyroid-and 2 rem to the whole body at the exclusion-area boundary and smaller within the ow population zone (LPZ) (SER,--13.6.1).
This analysis apparently.as-sumes that freshly discharged fuel assemblies will not be stored in an upper rack prsition directly over another freshly discharged fuel assembly.
The stated consequences are within the guidelines of 10 CFR Part 100.
Even if a freshly discharged assembly is stored over another such assembly, the offsite consequences are said to be within the ex-posure guidelines of 10 CFR Part 100, although no exposures are set forth.
Given these circumstances:
3.
Even though the maximum consequences of an accident (involving either a freshly discharged element stored over another such element, or a freshly discharged element stored over an older element) fall within the guidelines of 10 CFR Part 100, thev do not appnar to fall within the guidelines of EPA's Protective Action Guides (PAG).
These 116.2 624
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guides recwrend evacuation or other protective action where the ex-posure to the individual is 1-5 rem whole body and 5-25 ren thyroid.
See EPA " Manual of Protective Action Guides and Protective Actions for Nuclear Incidents," dated September,1975 (EPA-520/1-75-001),
Tables 2.1 and 2.2.
a.
In the event of a maximum fuel handling accident, are these EPA prescribed levels reached at any point outside the LPZ?
b.
If so, does the Applicant's emergency plan contemplate tak-ing of protective Actions to preclude exposures at these levels?
c.
In any event, does the Applicant's emergency plan provide adequate means to preclude persons in the exclusion area-or
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LPZ from being exposed to those levels'of radiation?-
Response
-(B.3.a) The potential consegences of the fuel handling accident with the postulated damage to tne described three freshly discharged fuel assemblies at the Low Population Zone (LPZ) boundary ae 4.2 Rom thyroid and 0.052 Rem whole body.
These maximum exposures and more realistically calculated exposures, are less than.the PAGs exposure guidelines. Therefore, in the event of a maximum fuel handling accident at LACBWR, the EPA prescribed exposure guideline levels are not reached at any point outside the LPZ.
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. (0.3.
The potential consequences of the fuel handling accident with b&c) the postulated damage to three freshly discharged fuel assemblies at the o' uter boundary of the Exclusion Area are 243 Rem Thyroid aid 3 Rem whole body for a 2-hour exposure.
The two-hour exposure through-out the Low Population Zone would be less.
An exposure of 243 Rem thyroid, if one takes no protective actions, is above the EPA Pro-tective Action Guides, and both the applicant's Emergency Plan and the Emergency Plan of offsite authorities specify protective measures
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in such a situation.
The licensee's Emergency Plan requires notifica-tion to offsite authorities if the projected exposure is calculated to be in excess of 1 Rem whole body or 5 Rem thyroid.
The licensee's Emergency Plan references the State of Wisconsin Peacetime Radiological Emergency Response Plan which recommends evacuation if an anticipated exposure is estimated to be in excess of 1 Rem whole body or 5 Rem thyroid. The licensee's plan specifies that the shift supervisor, acting as the Emergency Control Director, will notify the Vernon
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Sfieriff's Department directly to implement evacuation.
Consequently, we conclude that the coordination between the licensee's plan and the
plans 'of offsite authorities provide adequate means to protect the.
health and safety of the public in the event of an accident.
C.
It appears from Section 3.6 of the SER-that offsite doses for the fuel handling accident were calculated assuming that the containment building i16.2 326
, is not isolated.
Is this the case?
If so, how r.uch would the offsite doses be reduced if the containtr.ent building were isolated (1) at the time of the accident, and (2) as soon thereafter as practically achiev-able.
Please discuss whether containment building isolation should be required during fuel handling.
Response
The potential consequences of postulated fuel handling accidents at LACBWPs given in Section 3.6 of the Safety Evaluation for the proposed pool modification dated July 1979 are based on the containment build-ing n_ot being isolated.
If the containnent building were isolated at the time of the accident, there would probably be no release offsite.
The licensee does have a radiation monitor installed which will cause
' isolation of the containns ventilation as rapidly as practically achievable.
The staff recognizes this feature may provide a substantial reduction in offsite doses.
The monitor, which initiates containment ventilation isolation, is in place and by lechnical Specifications checked to provide assurance it will perform its intended function.
Because the potential consequences of damaging all the fuel pins in the equivalent of one freshly discharged fuel assembly, the most 1.ike,1y
' of the events analyzed, are appropriately within the exposure guide-lines of 10 CFR Part 100, the potential consequences of the fuel handling accidents at LACBWR are acceptable. Therefore, the con-
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tainment building ventilation isolation is not required during fuel handling opera t. ions.
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7-Isolating the containn.ent before the accidant will increase the occupational exposure to the workers in contairaent.
Keeping the containment building ventilation isolated during refueling will result in significantly higher concentrations of radioactivity in the air and higher occupational exposure to the workers in con tainment.
For this reason a Technical Specification requiring containment isolation during refueling has not been required.
D.
The SER discusses the cask drcp accident in terms of a previous Staff analysis of the existing spent fuel pool design.
Please pro-
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vide that portion of the October 22, 1975 SER describing that analy-sis and its results.
Do the consequences of this accident result in exposure levels requiring protective action under the EPA Guidelines referenced in question B?
Response
The potential consequences of the posutlated cask drop accident, given in the staff's safety evaluation are the maximum exposures to an individual at the Exclusion Area Boundary and the Low Popu-
' ' '1ation Zone boundary. These exposures are not what we would expect from such an accident at LACEWR.
Based on the extrapolation of,._
realistically calculated radiological consequences of spent fuel handling accidents at LACBWR given in Table 7.1-2 of the Draft Environmental Statement (DES) dated June 1976 for LACBWR, we esti-rute the expt.cted radiolcgical consequer.ces of a con drop acci-dont at LACBWR are 0.5 of 10 CFR Part 20 limits (0.75 Rem to 1162 M8
, thyroid) at site boundary and en estimated less than 25 man-ren to the population out to 50 miles.
These exp sures would not require protective action under the EPA Protective Action Guidelines (PAG).
The realistic assessment of radiological consequences for the cask drop accident is obtained by extrapolation of the expected conse-quences for a fuel handling accident given in Table 7.1-2 of the LACCWR DES by multiplying by 10 and 24 to account for all the fuel pins in each fuel assembly being damaged and account for all the fuel assemblies in the last refueling, respectively.
It is only the spent fuel from the last refueling that will contribute to the radiological consequences of the cask drop accident.
Enclosed is the October 22, 1975 Safety Evaluation describing the case drop accident and its results.
F.
The Staff's SER indicates generally tha't ALARA principles will be followed with respect to the exposure of workers (see SER, 53.7.2),
but no details of implementation are provided.
Indeed, the achieve-ment of Pa.rt 20 levels appears to be considered as equivalent of ALARA
~(see Staff's motion for summary disposition, p. 19).
The proposed amendment would increase occupational exposures in several ways, but information is not supplied which could be used to evaluate whether the resultant exposures are ALARA Specifically:
1.
If remote operations are used (Applicant's Plan J), the oc-cupational exposure is said to be about 16 man-rem; but, if remote i16.2 329
, operations are not used (Plo. A), the occupational exposures are said to be about 23 man-rcm.
The only justification for using plan A ap-pears, to be the lesser nut..ber of man-hours needed to complete the job.
See Applicant's letter dated October 26, 1978 (respc:se to question 3).
Is this standard consistent with ALARA requirements? What other cri-teria, if any, are to be used to ascertain whether remote or manual operations are to be employed?-
What cost differences, if any, exist?
2.
What occupational exposure differences, if any, would result from maintaining the water level in the pool at a 700' elevat.an rather than at level 16' above the storage racks? What cost differ.
ence, if any, attend each of these methods of operation?
3.
What criteria have been used to ascertain whether the in-crease of total plant maintenance exposures from increased filter changes and resin volumes and intensities occasioned by a larger, amount of spent fuel storage meet ALARA (not Part 20) standards?
Response
(F.1) As indicated by the October 26, 1978 letter referenced in the Board's question, plan B (remote operations) is the Applicant's preferred plan.
In the event this plan is not feasible either mechanically or eccnomically, plan A (divers) will be used.
The Applicant has not yet quantified the cost / benefit analysis of the alternative plans, since that will not be possible until the process of rec ate disa sembly is a tt.
%d.
Tho Aanlic:nt will r
- the final dccision as to choice of nethod taking into account the 116.?
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, alternate nethods, costs, and exposures so that the ALARA conceptr.
are achieved.
Presently DOP, policy is to require licensees that apply. for spent fuel pool n:odifications to p'esent the staff with alternative nethods of perfonning tasks to confina his comnittrent to ALAPA principles.
In judging whether a job is being perfonned with ALAPA exposure, we require a detailed breakdown of each phase of an operation uith man-rem exposure.
For Lacrosse, Plan A considers using divers to disconnect the old storage racks to allow the plant to get back to power with the least delay.
Plan B would disconnect the old racks remotely f rom the pool surface. Although Plan B would require less occupational exposure, it may involve added delay in plant start-up and consequent costs for replacement power for about seven days at $30,000 per day..(See EIA page 19).
The licensee will, therefore, re-evaluate the reduction in occupational exposure using Plan B against any delay in plant re-fueling and confirm that diver dose estinate is realistic before a final decision is made.
(F.2) The technical specifications for Lacrosse specifies that during normal operations, the water level over the SFP elec'ents shall be no less than 16'.
At this depth of water, the dose rate from the SFP elementa is about 1 mr/hr at the operating level of the SFP area.
The radio-nuclide concentrations in the SFP water will, however, be an important,
centributor t.o drse rne in ti e;mra t in t rea of ue J L'
'!ith the i 162
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, pool filled to elevation 700', the dose rate from the.tored fuel elements will t;e several orders of raagnitude lower due to the additional water shield.
However, at this elevation, the dose rate from the radionuclide concentrations in the SFP water is about 5 mrem /hr and is the major contributor to the dose rate in the SFP area. The reason for keeping the water level below the 700' elevation during normal operations (i.e., during refueling, the SFP water is at the 700' icvel) is to reduce potential leaks from the pool.
We do not know the cost difference, but believe it to be insignificant.
(F.1) The staff has reviewed the licensee's plan for routine changes of the SFP purification system filters and demineralizer resins. As a result of the modification of storage of spent fuel from 134 to 440 fuel assemblies, the licensee anticipates the frequency of change of filters and resin may double towards the end of storage capacity. This would increase the occupational exposure from about 1.5 man-rem / year to about 3 man-rem / year.
The effect of this more frequent change will be to reduce the radionuclide concent"tions in the SFP water and thus decrease the background radiation levels in de SFP area.
At the same time, the filters and demineralizer will become increasingly con-taminated and therefore may increase occupational exposures during maintenance operations.
Therefore, by changing the filters more frequently, present procedures can be used so that present dose rate levels are unchanged for each filter change.
This increase of 1.5 non-ron will ronrosant lase. tMn 1~ of the total anmn1 occupational exposure for the plant and would, therefare, represent a negligi.ble 116.! B2
, impact on occupat.icnal exposure rece!,cd t;y workers and is considered to be ALARA.
I have read the foregoing affidavit and swear that it is true and correct to the best of my knowledge and belief.
, g ( i ' 1's, t.J, j iC t ' v. ' \\,)
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Jack fi. Danchew
.j Subscribed and sworn to before me this
' day of
._, 1979
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