ML18079B044
| ML18079B044 | |
| Person / Time | |
|---|---|
| Site: | Salem, Crane  |
| Issue date: | 08/24/1979 |
| From: | Wetterhahn M CONNER, MOORE & CORBER, Public Service Enterprise Group |
| To: | |
| Shared Package | |
| ML18079B045 | List: |
| References | |
| NUDOCS 7910050014 | |
| Download: ML18079B044 (34) | |
Text
UNITED STATES OF A~IBRICA NUCLEAR REGULATORY COM.MISSION Before the Atomic Safety and Licensing Board In the Matter of
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PUBLIC SERVICE ELECTRIC AND GAS )
CO.MPANY, et al.
)
(Salem Nuclear Generating Station, Unit 1)
)
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Docket No. 50-272 (Proposed Issuance of Amendment to Facility Operating License No. DPR-70)
LICENSEE'S RESPONSE TO THE ATOMIC SAFETY AND LICENSING BOARD'S QUESTION 4 Introduction During the latest session of the evidentiary hearing in the captioned proceeding, the Atomic Safety and Licensing Board posed the following question:
The Proposed Annex to Appendix D, 10 CFR Part 50, appears to define a Class 9 accident as a sequence of failures which are more se-vere than those which the safety features of the plant are designed to prevent.
The sequence of failures at Three Mile Island produced a breach of the containment and a release oft radioactive material 1/ which could not be prevented by the safety features.
Was the occurrence at Three Mile Island therefore a Class 9 accident?
Was the risk to health and safety and the environment "remote in probability," or "extremely low" at Three Mile Island, as those terms are used in the Annex?
_]_/
Licensee, Public Service Electric & Gas Company, et al.,
hereby submits its response to the Board's question.
__ll See Tr. 925 for the correction by Mr. Shon.
_]_/
Tr. 922-3.
For ease of reference, this question has*
been designated Question 4.
7~ r oos6 o tel-
l 2
To address the Board's question fully, it is necessary to examine the methodology that the Cormnission has mandated for consideration of accidents in the course of the licensing process, from both safety and environmental viewpoints, and apply this methodology to the question posed by the Board.
Part 100, Title 10, Code of Federal Regulations "Reactor Site Criteria;" governs the siting and design, from a radio-logical safety viewpoint, of light-water-reactors such as the Three Mile Island Nuclear Station, Unit 2*, and the Salem Generating Station, Unit 1.
From an environmental viewpoint,
_l/
the proposed annex to 10 C.F.R. Part 50, Appendix D, is controlling as the nature and extent of Commission review.
We address these two regulations seriatim.
Safety Review Under the Atomic Energy Act Since 1962, 10 C.F.R. Part 100 has been the basis for reviewing the site suitability and the design for light-water-reactors.
Under Part 100, the major focus for design by a utility and review by the NRC is the identification of "major accident[s], hypothesized for purposes of site analysis or postulated from consideration of possible accidental events, that would result in hazards not exceeded by those
__!/
from any e.ccident considered credible.
11 Under Part 100, a key element regarding the determination of site suitability
~/ See 36 Fed. Reg. 22851 (December 1, 1971).
_!I 10 C.F.R. Part 100, n.l [emphasis supplied]
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from a radiological safety standpoint is whether the cal-culated dose consequences from the postulated accidents at 5/
the outer.boundaries of an "exclusion area"- and "low 6/
population zone ('L.P.Z. ')
11-are within specified dose
_J_/
guideline values.
Part 100 affects reactor design in that engineered safety features may be added or modified during the course of NRC review to further mitigate the conse-quences of accidents thereby making a site acceptable or by permitting the reduction of the size of the exclusion area and/or L.P.Z.
Part 100 notes the existence of Technical Information Document 14844 (TID 14844) which contains a procedural method and sample calculations which form the basis for the calculations required by 10 C.F.R. §100.11.
~/ 10 C.F.R. §100.3(a).
___§_/
10 C.F.R. §100.3(b).
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The guideline values are set forth in 10 C.F.R. §100.11:
(1)
An exclusion area of such size that an individual located at any point on its boundary for two hours immediately follow-ing onset of the postulated fission product release would not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
[footnotes omitted]
(2)
A low population zone of such size that an individual located at any point on its outer boundary who is exposed to the radioactive cloud resulting from the pos-tulated fission product release (during the entire period of its passage) would not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thvroid from iodine exposure.
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EnvirOThuental Reviews Under NEPA For major Federal actions significantly affecting the quality of the human envirolliuent, the National Environmental 8/
Policy Act~ requires a discussion of the environmental im-pact of the proposed action.
The NRC, and its predecessor, the Atomic Energy Corrunission, have considered during the course of their environmental reviews not only the impact of routine expected releases, but also the radiological en-vironrnental impacts associated with a spectrum of accident situations ranging from trivial to severe.
The Commission regulation which specifies the manner of treatment of ac-cidents in environmental impact statements for land-based light-water nuclear power plants, is the proposed Annex to 10 C.F.R. Part 50, Appendix D.
This regulation, its implemen-tation since it was announced on December 1, 1971 (36 Fed.
Reg. 22851), and its application and construction by the Commission and courts, were extensively discussed in "Licensee's Response to NRC Staff Objection to Board Question and Motion for Extension of Time.
." dated June 18, 1979, which is incorporated by reference herein.
With respect to the classes of accidents discussed within the Annex, Class 9 is the most severe.
With respect to Class 9 accidents, the Annex states:
The occurrences in Class 9 involve sequences of postulated successive
_§/
42 u.s.c. §4321 et seq.
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failures more severe than those pos-tulated for the design basis for protective systems and engineered safety features.
Their consequences could be severe.
However, the proba-bility of their occurrence is so small that their environmental risk is ex-tremely low.
Defense in depth (multiple physical barriers), quality assurance for design, manufacture, and operation, continued surveillance and testing, and conservative design are all applied to provide and maintain the required high degree of assurance that potential acci-dents in this class are, and will remain, sufficiently remote in probability that the environmental risk is extremely low.
For these reasons, it is not necessary to discuss such events in applicants 1 Environmental Reports.
Thus, the Commission has stated that Class 9 accidents need not be considered because "the environmental risk is extremely low."
From the above quoted passage, it is clear that the CoITLrnission has defined the term "environmental risk" as product of the probability of occurrence and the conse-quences of an event.
In this respect, it is essential to note that since Appendix D (and now Part 51) is seeking to address the potential environmental impact from the con-struction and operation of nuclear power plants, its focus is on the consequences of an event.
For example, there are many "successive failures" which produce trivial results.
Merely because there have been "successive failures" cer-tainly does not indicate that a Class 9 accident has oc-curred.
From the purpose of the Annex and its organization, it is clear that one must look to the consequences to deter-mine the existence of a Class 9 accident.
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Implementation of Part 100 and the Annex to 10 C.F.R. Part 50, Appendix D Included within the scope of the Commission's review of the Three Mile Island, Unit 2, facility was a comparison of the calculated dose from design basis accidents to the guideline values contained in 10 C.F.R. Part 100 and a finding that the requirement of 10 C.F.R. Part 100 were met.
Moreover, the Commission published an environmental impact statement, which included a discussion of various classes of accidents and the predicted radiological consequences in accordance with the Annex to Appendix D discussed above.
With regard to a demonstration of compliance with 10 C.F.R. Part 100, the Staff evaluated the potential offsite doses due to design accidents which included a loss-of-coolant accident, a fuel handling accident, steam line break
_Jj and gas decay tank rupture.
The controlling dose was produced by the loss of coolant accident; the two-hour whole body dose at the exclusion area was calculated to be 8.2 rem
!Q_/
with the thyroid dose being 280 rem.
It is important to note that of the accidents whose calculated radiological consequences are set forth in the SER, a number do not assume functioning of the containment.
Hypothesized acci-dents would take place outside of containment and therefore 2_1 10/
See Appendix 1.
Supplement No. l to the Safety Evaluation Report, TMI-2, dated March 11, 1977.
It must be noted that these doses are for only the first two hours and not the course of an accident.
The dose at the edge of the low population zone is 108 rem thyroid and 2.1 mrem whole body.
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no mitigation credit for the containment is taken.
For accidents which occur in the containment, a leakage rate of the containment is assumed, with separate consideration given to pathways which might bypass the containment struc-ture, such as valve leakage.
With regar~ to the environmental consequences of pos-tulated accidents, these are considered in §VI of the Final Environmental Statement related to operation of the Three
.Mile Island Nuclear Station, Units 1 and 2, dated December 1972 (Attachffient 2, hereto).
Table 20 SQrnmarizes the en-vironmental consequences of accidents considered.
The so-called Class 8 accidents, i.e., those accident initiation events considered in design basis evaluation in the Safety Evaluation Report, are of interest here.
These events included small and large loss of coolant accidents, a rod ejection accident, and small and large breaks in the steamlines outside of containment.
As far as doses are concerned, the large loss-of-coolant accident is controlling, with a cal-11/
culated dose of 600 millirem~ at the site boundary.
The Three Mile Island Event As discussed in "Licensee's Response to Licensing Board's Question 1 and Part 1 of Question 3 Relating to Impact of a Three Mile Island Type Iricident on the Salem 11/
The chart shows the dose as an estimated fraction of the 10 C.F.R. Part 20 lir.'li t at the site boundary.
Foot-note 1 states that the dose number represents the cal-culated fraction of a whole body dose of 500 mrem or the equivalent dose to an organ.
Multiplying 1.2 by 500 mrem gives 600 millirem.
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e e
Unit 1 Spent Fuel Pool" (Tr. following 1264), the following sequence of events is postulated to have occurred at TMI-2:
[T]he accident apparently resulted in the release of radioactive fission pro-ducts contained in the fuel rod gaps (void spaces between the uranium dioxide fuel pellets and the zircaloy cladding) in the reactor core due to clad failure.
In addition, radioactive noble gases in the fuel pellet were apparently released from the core resulting in a total estimated release of approximately 30% of the core radioactive gases.
The activity was re-leased to the primary reactor cooling water, a portion of which was spilled onto the reactor containment floor.
Some of the gaseous activity in the released reactor coolant, particularly noble gases, became airborne in the containment structure.
The spilled reactor coolant collected in the containment sump, where a portion was pumped into the auxiliary building liquid waste storage tanks.
Some of this water spilled onto the auxiliary building floor, resulting in the radiation levels measured in the auxiliary building and substantially all of the radiation levels in the surrounding environs.
The latest information known to the Licensee is that as a result of the accident at TMI-2, the maximum measured dose at an offsite location was less than 83 rnillirern, at a loca-tion 1200 meters north-northeast of the site with a calculated maximw~ thyroid inhalation dose of 10 millirem.
Based upon actual measurements of cow's milk, the maxinw~ ingestion dose to the thyroid is conservatively calculated to be 2.3 12/
millirem.-
QI Second Interim Report on the Three Mile Island i\\"uclear Station, Unit 2 (TMI-2) Accident (Executive Su.rnmary, Radiological Monitoring) dated June 15, 1979, and trans-mitted by Metropolitan Edison to NRC Region I on June 18, 1979.
I
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The Board's question states that the sequence of fail-ures at Three Mile Island "produced a breach of the con-tainment and a release of radioactive material which would not be prevented by the safety features."
This premise re-quires exploration.
Initially, at TMI-2 there was not a breach of the contaiTu~ent in the same sense as usually used in discussing Class 9 accidents, i.e., an irreversible breach such as a hole in the large reinforced concrete, steel lined building which occurs as a r~sult of over-pressure, as a result of some explosion due to some action of the primary system or some other mechanism which is the result of the initiating accident.
The pathway at T.MI-2 resulted from, inter alia, a line which, in the system as designed, was not called to be isolated in a particular set of circumstances; from all reports, the containment sump system did shut down and did isolate when called on to do so by the operators.
Aside from the release of water into the auxiliary building which was controlled during the course of the accident, the containment system, including the isolation valves, appeared to have performed its function satisfac-torily.
As previously discussed, the NRC's methodology for accident assessment contemplates that there may be releases which bypass the containment or leak out of the
.I-.
con.... ainmen -c.
l through the structure or through barriers such as isolation
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valves.
An example previously discussed is the PWR steam line break accident which is hypothesized to occur outside containment, with a leakage path through the steam generator assumed.
Thus it may be seen that pathways do exist for bypass of the containment; if such bypass is termed in the broadest sense a "breach," it is accepted and taken into account in NRC dose calculations.
Perhaps, conceptually, with regard to light-water-reactors, the clearest example of an accident which bypasses containment is a break in an instrument line in a boiling water reactor immediately outside containment.
The break is considered non-isolable until the accident has run its course.
The consequences of the accident are calculated in accordance with the procedures established by the NRC; however, merely because the containment has been bypassed, and a release of radioactivity which could not be prevented by the installed safety features occurs, this accident is certainly not a Class 9 accident.
The consequences of the accident must be examined in order to make a judgment with regard to the existence of a Class 9 accident and its as-sociated risk.
Thus, it may be seen that the consequences of accidents must be examined as an element in determining their class.
As has been previously indicated, the maximum dose which would have been experienced offsite at TMI-2 by a hypothe-
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e e
tical individual was on the order of 83 millirem whole body, and on the order of 10 millirem thyroid inhalation dose.
With regard to thyroid dose, the calculated conse-quences of the TMI-2 occurrence is a factor of 20,000 less than calculated by the NRC with regard to the site 13/
suitability guidelines contained in 10 C.F.R. Part 100.~
Even if the calculated dose at TMI-2 is compared with the radiological consequences set forth in the Final Environ-mental Statement for Three Mile Island, it is small.
The large loss-of-coolant accident dose was calculated to cause a dose of 600 millirem, with a small loss of coolant acci-14/
dent computed to be approximately 80 millirem.~ Thus, the calculated doses to a hypothetical individual based on measured data at TMI as a result of the recent accident are lower or of the same order of magnitude as certain of the calculated doses presented in the Final Environmental Impact Statement for that facility.
On August 22, 1979, the Licensee received a copy of Chauncey Kepford's technical report in response to the Board Question #4.
While there has been only a very limited opportunity to review Dr. Kepford's response, the initial conclusion is that there are two serious errors relating to 13/
14/
This factor is conservative for the additional reason that a two-hour dose is being compared to a course-of-accident dose.
The 80 millirem dose was calculated in a manner similar to that presented in footnote l~
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the projected dose calculations presented therein which destroy any validity it might have.
The 280 rem whole body dose calculated on page 4 of that document was ap-parently based on Three Mile Island FSAR meteorological factors associated with the most unfavorable two hour time period at the 610 meter exclusion radius.
Actual rneteoro-logical factors during the Three Mile Island incident were significantly lower (i.e., better dispersion) than the con-servative FSAR asslli~ptions.
Furthermore, the release of XE-133 and I-131 occurred over a period of several days.
For this situation, meteorological factors associated with a several day period at the 3200 meter Low Population Zone distance must be used.
It is not appropriate to calculate a dose at the exclusion radius for an accidental release that occurs over a several day period since members of the public within the low population zone would be evacuated if potential exposures were projected to exceed even a few rem.
The meteorological dispersion factor at the low population zone distance for the one to four day time interval after a design basis accident is 2.sx10-S sec/m3, a factor of 44
-3 I 3 lower than the l.lXlO sec m value at the exclusion radius for a 0-2 hour time interval.
The 280 rem whole body dose is therefore a fictitious number as a result of the use of inappropriate meteorological factors.
It is beyond belief that individuals received a whole body dose approaching 280 rem over a period of days without anyone detecting it.
To l
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the contrary, in the July 23, 1979 Federal Register (44 Fed. Reg. 43128-31) "Financial Protection Requirements and Indemnity Agreements; Section 82 - Procedures, 11 the Cornmis-sion indicated that the approximate upper limit on whole body dose to a person in a populated area was calculated to be 100 millirem.
Dr. Kepford's calculation of 63,000 person-rem (page 6 of his report) also appears to be based on inappropriate assumptions.
Initially, he refers to a Figure 1 on page 7 of his report.
However, no copy of "Figure l" was attached to the copy of the report received by Licensee.
- Thus,
- .l-l L.
is not possible to follow his methodology completely.
.More-over, on page 8, reference is made to ~dose of 1 mrem at 50 miles by use of a "straight line" approximation.
Such a linear interpolation appears to be technically incorrect and inconsistent with any standard references relating to atmospheric dispersion of any airborne material.
On page 36 of NUREG-0558, a statement is made regarding a DOE analysis that suggests that the dose decreased with distance at a much faster rate (an exponent of -2) than assu..med in the actual calculations.
Therefore, the approximately 3500 person-rem dose from the NUREG-0558 doclli~ent referred to on page 6 of Dr. Kepford's report is conservative and overestimates actual population exposure, contrary to Dr. Kepford's un-supported contention that the calculation was unconservative.
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Using the definition of Class 9 accidents as presented in the Annex to Appendix D, the occurrence at TMI-2 was not a Class 9 accident in that the consequences are no more 15/
severe than Class 8 accidents which have been analyzed.~
As discussed previously, as used by the NRC, in the Annex to Appendix D, risk is the product of the probability and the consequences.
Since the consequences are relatively low, the risk is therefore correspondingly relatively low.
Moreover, Appendix D to 10 C.F.R. Part 50 adds the following proviso to the consideration of accidents:
Furthermore, it is not necessary to take into account those Class 8 accidents for which the applicant can demonstrate that the probability has been reduced and thereby the calculated risk to the envi-ronment made equivalent to that which might be hypothesized for a Class 9 event.
Assuming, arguendo, that the sequences of the TMI event would have had to be considered during the TMI-2 licensing process, it would not necessarily follow that the TMI-2 scenario would have to be considered in the present pro-ceeding for Salem Unit 1.
During the course of the present proceeding, it was made clear that certain design differences exist and design changes have been made which preclude a TMI-2-type event 16/
from occurring at Salem Unit l.~ Therefore, pursuant to
_!21 It should also be noted that in the past a number of incidents have occurred at operating nuclear reactors where more than one failure has been involved.
- However, at no time has the NRC indicated that a Class 9 accident had occurred.
Again, this is indicative of the fact that the consequences must be examined to make this determination.
l§_/
Tr. 1281-5.
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the above quoted section of Appendix D, the probability of the event has been reduced to such an extent that it need not be ~aken into account.
This fact, together with the low level of consequences, make the risk quite small.
However, assuming, arguendo, that the TMI occurrence was nominally referred to as a Class 9 accident, Licensee submits that this does not in and of itself open the flood-gate to consideration of all possible accidents which could be hypothesized for this facility.
Without attempting to fully explore the limits on accidents which may have to be explored in the future for facilities seeking construction permits or operating licenses, the range of accidents which are initiated by a core meltdown is not appropriate for 17/
consideration in this (or any other) proceeding.-
From measurements taken of the core fuel temperatures, it has been concluded that temperatures sufficient to cause core 18/
melting were probably not reached.-
Moreover, the hydrogen explosion problem which held the headlines in the days after the incident, has subsequently been minimized.
There is absolutely no basis to extrapolate the TMI-2 accident into an event which would require consideration of the consequences of accidents associated with core meltdown.
17/
In NUREG-0440, Liquid P.athway Generic Study, Impacts of Accidental Radioactive Releases to the Hydrosphere from Floating and Land-Based Nuclear Power Plants, dated February 1978, it is clear that accidents which are beyond the design bases are those associated with the so-called core melt accident.
Similarly, the Atomic Safety and Li-censing Board spoke of only core melt accidents as fittins the Commission's description of a Class 9 accident.
Q-f-f-shore Power Systems (Floating Nuclear Power Plants), ALA.D-489, 8 :N"RC 194, 209, 211 (1978).
l....§_/
NUREG-0557 at 2-1.
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e e
This.is particularly true in the present proceeding.
It must be constantly borne in mind that the general question of Class 9 accidents is peripheral to the issues before this Board, the proposed expansion of the spent fuel pool capacity for a single reactor.
In addition, the matter arose as part of a Board inquiry, rather than as part of the consideration of an admitted contention.
While a decision to consider any type of a Class 9 accident would undoubtedly cause extremely serious delays in this proceeding and may well require extended evidentiary sessions, these effects would be mild 19/
compared to its effect in other cases.-
The Licensee submits that absent direct and explicit guidance from the Commission, this Board should not attempt to consider Class 9 accidents in this proceeding.
On July 2, 1979, various atomic safety boards, including the Licensing Board in this proceeding, were notified of the fact of certain TMI release estimates and the "possibility that the calculations used by the Staff to judge the suita-
~/
bility of sites may be substantially nonconservative."
The memorandum noted that an additional memorandum was attached which provided the basis for a conclusion "that the
'potential hazard' from the TMI accident source did not
.!2._I It should be noted that other licensing boards have, subsequent to the Three Mile Island accident, decided that Class 9 accidents should *not be considered.
- See, for example, Boston Edison Company (Pilgrim Nuclear--
Generating Station, Unit 2), Docket No. 50-471 CP, Rulings on Commonweal th.Motions to Enlarge Time for Filing of Testimony and to Defer Evidentiary Sessions, dated August 9, 1979, at 1, n.l.
~/ Memorandum to E.S. Christenbury, Chief Hearing Counsel, OELD, from Steven A. Varga, Acting Assistant Director for Light Water Reactors, Division of Project Management, dated June 27, 1979.
17 -
-1/
exceed that used in the licensing process.
11 The memoran-dwn stated the Staff position that it had 11 not concluded as to the materiality or relevancy of this recommendation for 22/
any particular proceeding."-
As discussed below, this notification does not affect the above discussion regarding Class 9 accidents in the proceeding.
Licensee submits that the Barrett memorandum noted in 23/
the June 27, 1979 memorandum discussed above-explains the fact that while the number of curi~s of Xe-133 was greater than would have been predicted by the use of TID-14844, the "potential hazard 11 from the TMI accident source does not exceed that used in the licensing process.
As pointed out in the Barret memorandum, there are other factors that must be taken into account, namely the energy of the isotope and the mix assumed.
Thus, as demonstrated by the ultimate risk of "hazard," the calculated dose received, the potential hazard as previously reported in the Safety Evaluation Report for TMI-2 (Attachment 2) is not exceeded and the 24/
analyses remain valid.~
21/
22/
~/
3.4.
_-/
Id.
Id.
Memorandwn to D. Bunch, Director, Program Support Staff, Nuclear Reactor Regulation, from L. Barrett, Section Leader, Environmental Evaluation Branch, DOR.
That is not to say that for the long term, the NRC will not modify the source terms used to analyze acci-dents to take into account the occurrence at TMI-2.
It goes without saying that the NRC will utilize the best available information in the licensing process.
It has already begun to do so.
That does not change the fact, however, that the "potential hazard" as analyzed in the SER is not exceeded by the TMI-2 incident.
Conclusion For the above stated reasons, the risk to health and safety and the environment was remote in probability and extremely low as those terms are used in the AI1nex to Appendix D, 10 C.F.R. Part 50 and the TMI occurrence was not a Class 9 accident.
For these reasons and the reasons stated in Licensee's Response to NRC Staff Objection to Board Question and Motion of an Extension of Time to File Response to Board Question Relating to Class 9 Accidents dated June 18, 1979, no consideration of a "meltdown" and 11 explosion 11 should be given in this proceeding.
Any other course would effect an unwarranted radical departure from Commission precedents.
By deciding to consider so-called Class 9 accidents, a sweeping precedent would be established by this Board radically affecting all pending cases (and perhaps those for which licenses have already been issued) in a matter which is really peripheral to the resolution of the issues in this proceeding.
As originally requested in Section IV of the June 18, 1978 pleading, should the Board decide that consideration of Class 9 accidents is necessary, the Board should certify the matter or refer it to the Commission for its determination.
Respectfully submitted, CONNER, MOORE & CORBER etterhahn for the Applicants August 24, 1979
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VI.
ENVIRONMENTAL IKPACT OF POSTULATED ACCIDENTS A.
PLANT ACCIDENTS A high degree of protection against the occurrence of postulated accidents in the Three Mile Isla.'1d Nuclear Station, Unit 1 and Unit 2, is provided through*
correct design, manufacture, and operation-, and the quality assura..'1ce program used to establish the necessary high integrity of the reactor system as will be
.considered in the Co!II!Ili:ssion's Safety Evaluation for each.unit.
Deviations that rr:ay occur are ha..'1dled by protective systems to place and hold the plant in a safe condition.
Notw-ithstanding this, the conservative postulate is made that serious accidents might occur, in spite of the fact that they are extremely unlikely; and engineered safety features are installed to ::litigate the consequences of these postulated events.
rne probability of occurrence of accidents and the spectrum of their conse-quences to be considered from an environmental effects standpoint have been ana-lyzed using best estimates of probabilities and realistic fission product release and transport assun::ptions.
For site evaluation in t~e Staff safety review, extremely con~~=vative assumptions were used for the purpose of co~paring calcu-lated doses resulting from a hypothetical release of fission products from the fuel against the 10 CFR Part 100 siting guidelines.
The corrputed doses that
- would be received by' the population and environment from actual accidents would be significantly less than those that will be presented in the Staff Safety Evaluations.
rne Com.ruission issued guidance to Applicants on September 1, 1971, requiring the consideration of a spectrum of accidents with assumptions as realistic as the state of knariledge permits.
The Applicants' response is contained in "Environmental Report - Operating License Stage" for the Three Mile Island Nuclear Station, Unit l m:d Unit 2, dated Dece~~er 10, 1971~
The Applicants' report has been ev'lluated using the standard accident assumi?tions and
~~.;id<:.nce issued as a proposed Annex to Appendix D of 10 CF~
Part 50 by d1e Comrni:>sion on Dr:;cer:ib2r 1, 19 71.
Nine classes of pas tulated o.cci,foats anJ occurrences ranging in severity from trivial to very serious were.Ldentified by the Com.mission.
In general, accidents in the'* high potential consequence end of the spectrur:i. have a.low occurrence rate, ar1d those on the low potential consequence end have a higher occurrence rate.
The 2xau:ples selected by the Applicants for these classes are shown in Table 20.
Ihe exar:ples selected are reasonably hooogeneous in terms of probability with t*.;o exceptions.
It was considered to be more appropriate to classify (1) the
- ailure of the waste gas decay tank as an accident in Class 3 (Applicants use
- lass 8) and (2) the steam generator tube rupture as an accident in Class 5 VI-1
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VI-2 (Applicants use Class 8).
The following assumptions made by the Applicapts are questionable:
(1) no steam generator tube leaks prior to the steam generator tube rupture are considered, (2) the primary coolant activity is based on 0.1% failed fuel, and (3) the consequences of various releases are evaluated based on release rates applicable for specified times.
- However, the use of.alternative assumptions does not significantly affect overall environmental risks.
.......~~
The postulated occurrences in Cl.sss 9 involve failures more severe than those reauired to be considered for the design basis of protection systems and engineered safety features.
Tli.e:i.i: consequences could be severe.
How--
ever, the prob~bility of their occurrence is so small that their en~iron mental risk is extremely low.
Defense in depth (multiple physical barriers),
quality assurance for des{gn, manufacture, and operatio~, continued surveil-lance and.testing, and conservative design are all applied to provide and maintain the required ~igh degree of assurance that potential accidents in this class are, and will remain, sufficiently small. in probability that the environmental risk is extremely low.
Staff estimates of the dose which i:1ight be received by an assumed individual standing at the site boundary in the downwind direction, using the assuaptions.
in the proposed Annex to Appendix D, are presented in Table 20.
Estimates of the integrated exposure that might be delivered to the population within 50 railes of the site are also presented in Table 20.
The man~rem estinate was based on the projected population around the site for the year 2014.
The estimates pre-sented in Table 20 refer to a single unit.
To rigorously establish a realistic annual risk, the calculated doses in Table 20 would have to be multiplied by estit:1ated probabilities.
The events in Classes 1 and 2 represent occurrences i;.;hich are anticipated during plant operation and their consequences, which e.re very small, are considered within the framework of routine effluents fro:n the plant.
Except for a limited CJ.rr:our:t of fuel failures and sane stean generator. leakage, the events in Cl2sses 3 through 5 are not anticipated during.plant operation; but events of this type could occur sometime during the 40-year plant lifetime.
Accidents in Classes 6 and 7 and small accidents in Class ~ are of sinilar or lower probability than accidents in Classes 3 through 5 but are still possible.
The probability of occurrence of larg~ Class 8 accidents is very small.
Therefore, when the consequences indicated in Table 20 are weighed by probabilities, the environ-mental risk is very low.
The postulated occurrences in Cl~ss 9 involve sequences of successive failures more severe than those required to be considered in the design basis of protection syste~s and engineered safety features.
Their con-seque~ces could be severe.
However, the probability of their occurrence is so small that their environmental risk is extrer.:ely low.
Defense in depth (r.cultiple physical barriers), quality assurance for design, manufacture and operation,
VI-3 continued surveillance and testing, and conservative design all are applied to provide and maintain the required high degree of assurance that potential acci-dents in this class are, and will remain sufficiently small in probability that the environmental risk is extremely low.
Table 20 indicates that the realistically estimated radiological consequences of the postulated accidents would result in exposures of an assumed individual at the site boundary to concentrations of radioactive materials within or comparable to the Haximum. Permissible Concentrations (MPC) of Table II of 10 CFR Part 20.
The table also shows that the estimated integrated exposure of the population within 50 miles of the plant from each postulated accident would be orders of magnitude smaller than that from naturally occurring radioactivity, which corresponds to approximately 394,000 man-rem/yr based on a natural background level of 130 p:,.rem/yr.
When considered with the probability of occurrence, the ann~al potential radiation exposure of the population from all the postulated accidents is an even smaller fraction of the exposure from natural background radiation and, in fact, is well within naturally occurring variations in the natural background.
It is concluded from the results of the realistic analysis that the environmental risks due to postulated radiological accidents are exceedingly small.
B.
TRANSPORTATION ACCIDENTS*
- 1.
Ne~.; Fuel Under accident conditions other than accidental criticality, the pel-letized form of the nuclear fuel, its encapsulation, and the low specific activity of the fuel limit the radiological impact on the environment to negligible levels.
The packaging is designed to prevent critir.ality under normal and
~cvere accident conditions.
To release a number of fuel assemblies und~r conditions that could lead to accidental criticality would require severe damage or dest~uction of more than one package, which is unlikely to hap-
µen in other than an extreme.Ly severe accident.
TI1e probability that an accident could occur under conditions that could result in acci~ental criticality is e~tremely reractc.~ If critl~allty were to occur in transport, persons within a radius of about 100 feet from the accident might receive a serious exposure but beyond that distance, no detectable radiation effects would be likely.
Persons ~ithin a few feet of the accident could receive fatal or near-fatal exposures unless shielded by intervening material.
Although there would be no nuclear explosion, heat
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9 VI-4 TABLE 19 CLASS I FI CATION OF POSTULATED ACCIDENTS AND OCCUR.."R.ENCES AEC Descriotion Trivial Incidents Small Releases Outside Con-taini:;::ent
' Radwaste Systeci. Failure Fission Products to Prisary Sys te:n (BWR)
Fission Products to Primary_
and Secondary Systems (PWR)
Re.fueling Accidents Spent Fuel Handling Accident Accident Initiation Events Considered in Design Basis Evaluation in the Safety Ari.alysis Re po rt Hypothetical Sequences of Failures Hore Severe Tha.:.J.
Class 8 Aoolicant's Exa.r:mle(s)
None*
Spill in Sample Hood Inadvertent Release of Waste Gas Decay Tank Not applicable One day Operation with Primary System Leak to Reactor Bi_iilding Norr:ial Operation with Stean Generator Tube Leak and Release from Condenser Drop of Fuel Assembly or Drop of Heavy Object on Fuel Assembly Drop of Fuel Ass erclJ ly Uncompensated Operating Reactivity Changes Startup Accident Kod Withdrawal Acciden;:
Cold Water Accident Loss of Coola...<t Flow Accident Stuck-Out, Stuck-In, or Dropped Control Rod Accident Less of Electric Load Accident Steam Line.Failure Stea~ Line Le~~age Steam Generator Tube Failure Rod Ejection Accident Loss of Coolar1 t. Accident Waste Gas Tank Rupture None I l i
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Class 1.0 2.0 3.0 3.1 3.2 3.3 4.0 5.0 5.1
.5.2 5.3 VI-5 TABLE 20 SU}frf.ARY OF RADIOLOGICAL CONSEQUENCES OF POSTULATED ACCIDENTS (Single Unit Only)
Event Trivial incidents Small relecses outside containment Rad~aste system failures Equipment leakage or malfunction Release of waste gas storage tank contents Release of liquid waste storage tank contents Fission products to primary system (B'rlR)
Fission products to primary and secondary systems (PWR)
Fuel cladding defects &"1d steam ge~erator leaks O"ff-design tra..1sient3 the.t
- induce fuel f.2.ilure a~ove those expected and steam generator leak Steam generator tube ::rupture Estimated Fraction of 10 CFR Part 20 Limit at Site Boundary])
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Jj 0.073 0.29 0.003 N.A.
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.002 0.096 Estimated Dose to Population in SO Mile Radius, man-rem 10 40 0.47 N.A.
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VI-6 TABLE 20 (cont 1 d)
Class Event Estimated Fraction of 10 CFR Part 20 Limit at Site Boundarv_~/
6.0 Refueling accidents 6.1 6.2 Fuel bundle drop Heavy object drop onto fuel in core 7.0 Spent fuel handling accident 7.1 7.2 7.3 Fuel assembly drop in fuel storagi:" pool Heavy object drop onto fuel rack Fuel cask drop 8.0 Accident initiation events considered in design bas is evaluation in the safety analysis report 8.1 Loss-of-coolant accidents
- 8. l(a)
- 8. 2 (a)
- 8. 2 (b)
- 8. 3(a)
Small Break Large Break Break in instru2ent line fron priu:.ary system that penetrates the containment Rod ejection accident (PWR)
Rod drop accident (BWR)
Steamline breaks (PWR's-outside containment)
Sir.all Break Large Break 0.015 0.26 0.01 0.038 N.A.
0.16 1.2 N.A.
0.12 N.A.
<0.001
<0.001 EstiL;J.ated Dose to Population in 50 Mile Radius, mcn-rem 2.1 36
- 1. 3 5.3 N.A.
40 1000 N.A.
100 N.A.
<0.1 0.13
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VI-7 TABLE 20 (cont'd)
Class Event 8, 3(b)
Steamline breaks (BWR)
Estimated Fraction of 10 CFR Part 20 Limit at Site Boundaryl/
N. A.
Estimated Dose to Population in 50 Mile Radius, I:J.an-rem N.A.
1../ Represents the calculated fraction of a whole body dose of 500 mrem or the equivalent dose to fui org~1.
!:._/ These rele&ses will be comparable to the design objectives indicated in the proposed Appendix I to 10 CFR Part SO for routine effluents (i.e., 5 rnxem/yr to an individual frou either liquid or gaseous effluents),
VI-8
- enerated in the reaction would probably separate the fuel elecents so that
- he reaction would stop.
The reaction would not be expected to continue
- or more than a few seconds and normally would not recur.
Residual radiation
.evels due to induced radioactivity in the fuel elements night reach a few
- oentgeP£ per hour at 3 feet.
There.would be very little dispersion of radio.activ.e naterial.
- 2.
Irradiated Fuel Effects on the enviror....~ent from accidental releases of radioactive materials situation gases and during shipment of irradiated fuel have bee~ estimated for the
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(a) Leakage of contaninated coolant resulting fro~ improper closing of the cask is 1_Jossible as a result of human error, even though the shipper is required to follow specific procedures which include tests and exanination cf the closed container prior to each shipment.
Such an accident is highly unlikely during the 40-year life of the plant.
Leakage of liquid at a rate of 0.001 cc per second or about 80 drops/hour is about the snallest amount of leakage that can be detected by visual obse::::-vation of a large container.
If undetected leakage of contaminated liquid coolant were to occur, the amount would be so small that the individual exposure would not exceed a few mrem and only a very few people would receive such exposures.
(b) Release of gases and coolant is an extremely remote possibility.
In the improbable event that a cask is involved in an extremely severe accident such chat the cask c0ntairunent is breached and the cladding of the fuel assemblies penetrated, some of the coolant and some of the noble gases might be released from the cask.
In such an accident, the a.mount of radioactive ~aterial released
~ould be limited to the available fraction of the noble gases in the void spaces in the fuel pins and so~e fraction of the low level contamination in the coolant.
Persons would not be expected to remain near the accident due to the severe conditior:s T~'hich would be involved, including a l:Cajor fire.
If releases occurred, they would be expected to take place in a short period of tir:i.e.
Only a limited area would be affected.
Persons in the dor,,-nwind region and 'W'ithin 100 feet or so of the accident might receive doses as hig~ as a few hundred millirem.
Under average weather conditions, a few hundred square feet might be cont3::i.inated to the extent that it would require decontamination (that is, Range I contamination levels) according to the standards 1 of the Environ;:i.ental Protection Agency.
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VI-9
- 3.
Solid Radioactive Wastes It is highly unlikely that a shipment of solid radioactive ~aste will be involved in a severe accident during the 40-year life of the _plant.
If a shipment of low-level wc.ste (in dru.:ns) becomes involved in a severe 1
accident, some release of waste might occur but the speci~ic activity of the
~
waste will be so low that the exposure of personnel would not be expected to be significant.
Other solid radioactive wastes will be shipped in Type B packages.
The probability of release from a Type B package, in even a verJ severe accident, is sufficiently s'G:'.Eill that, considering the solid form of the waste and the very remote probability that a shipment of such waste would be involved in a very severe accident, the likelihood *of s*ignitics-i.t exposure would be extremely small.
In either case, spread of the conta~ination beyond the irn:::.~diate area is.u.11.likely and, although local cle.an-up might be required, no significant exposure to the general public would b-e e}:pected to result.
- 4.
Severity of Postulated Transportation Accidents The events postulated in this analysis are unlikely but possible.
Hore severe accidents than those analyzed can be postulated and their conse-quences could be severe.
Quality assurance for design, manufacture, and use of the packages, continued surveill~11.ce and testing of packages and transport conditions, a,..1d conservative desigci of packages ensure that the probability of accidents of this latter potential is sufficiently small that the environ-mental risk is extremely low.
For those reasons, mare severe accidents have not been included in the walysis.
S.
Alternatives to Normal Trar1sportation Procedures Alten1atives, such as special routing of shipments, providing escorts in separate vehicles, adding shielding to the containers, and constructing a fue: recovery and fabricacion plant on the site rather ~ha.a shipping fuel to and fron the station, hc:.ve been exacined by the Staff on a generic 'basis.
The irr:p::.ct on the environwent of transportation t:nder normal or postulated e.ccident conditions is. not considered to be suffici:=nt to justify the addi-tion;;:.:L effort required.to in:.?lecen t :m.y of the alti?rr:atives.
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VI-10 References For Section VI
- 1.
Federal Radiation Council Report No. 7 "Background Haterial for the Development of Radiation Protection Standards; Protective Action Guides for Strontium 89, Strontium 90 and Cesiur:i. 137." Hay 1965.
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UNITED STATES OF AMERICA NUCLEAR REGULATORY CO.MMISSION Before the Atomic Safety and Licensing Board In the Matter of PUBLIC SERVICE ELECTRIC Al~D GAS COMPANY, et al.
(Salem Nuclear Generating Station, Unit 1)
)
)
) Docket No. 50-272
)
(Proposed Issuance of
) Amendment to Facility
) Operating License
) No. DPR-70)
CERTIFICATE OF SERVICE I hereby certify that copies of 11 Licensee 1 s Response to the Atomic Safety and Licensing Board's Question 4,"
dated August 24, 1979, in the captioned matter, have been served upon the following by deposit in the United States mail, this 24th day of August, 1979:
Gary L. Milhollin, Esq.
Chairman, Atomic Safety and Licensing Board 1815 Jefferson Street Madison, Wisconsin 53711 Mr. Frederick J. Shon Member, Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Dr. James C. Lamb, III Member, Atomic Safety and Licensing Board Panel 313 Woodhaven Road Chapel Hill, N.C.
27514 Chairman, Atomic Safety and Licensing Appeal Board Panel U.S. Nuclear Regulatory Comrnission Washington, D.C.
20555 Chairman, Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Barry Smith, Esq.
Office of the Executive Legal Director U.S. Nuclear Regulatory Com."TTiission Washington, D.C.
20555 Richard Hluchan, Esq.
Deputy Attorney General Department of Law and Public Safety Environmental Protection Section 36 West State Street Trenton, N.J.
08625
2 -
Richard Fryling, Jr., Esq.
Assistant General Solicitor Public Service Electric
& Gas Company 80 Park Place Newark, N. J.
07101 Keith Onsdorff, Esq.
Assistant Deputy Public Advocate Department of the Public Advocate Division of Public Interest Advocacy Post Off ice Box 141 Trenton, N. J.
08601 Sandra T. Ayres, Esq.
Department of the Public Advocate 520 East State Street Trenton, N. J.
08625 Mr. Alfred C. Coleman, Jr.
Mrs. Eleanor G. Coleman 35 "K" Drive Pennsville, New Jersey 08070 Carl Valore, Jr., Esq.
Valore, McAllister, Aron
& Westmoreland Mainland Professional Plaza P. 0. Box 175 Northfield, N. J.
08225 Off ice of the Secretary Docketing and Service Section U.S. Nuclear Regulatory Commission Washington, D. C.
20555 June D. MacArtor, Esq.
Deputy Attorney General Tatnall Building, P. o. Box 1401 Dover, Delaware 19901 J. Wetterhahn
- J