ML20127A500
Text
L FEB 2 21983 MEMORANDUM FOR: William Dircks, Executive Director for Operations FROM: Harold R. Denton, Director, Office of Huclear Reactor Regulation
SUBJECT:
NUREG-0771 UPDATE .
The Accident Source Tem Program Plan that you transmitted to the Office Directors, dated December 17, 1932, contained a February 1933 milestcae '
for the issuance of an updated version of NUREG-0771, " Regulatory Imp et -
of Nuclear Reactor Accident Sotrree Tem Assumptions." Lead for prepara-tion of this document has been in NRR with assistance from RES. As a result of the subsequent femation of the Source Tem Program Office and progress on interim source tem revisions, we have reconsidered the scope and thrust of this document and have concluded that further effort to complete it as currently structured to include severe accident considera-tions is not warranted. The insights gained during our work gn this pro-posed docuinent will not be lost, however, since its principal author.
W. F; Pasedag, has now been detailed to the new Program Office.
In place of'.it, we' hav'e 'extraN5$! $aterial prepared for the 0771 Update -
that' addresses accident source tems used in current design basis accident licensing practice with the discussion particularly oriented toward the impact of possible changes. The enclosure, " Design Basis Radiological Accident Source Tems," is the result of this effort. This distillation
- of our licensing experience should provide useful input to the preparation of the staff report currently identified as item 15 in the Accident Source Tenn Program Plan on proposed changes to licensing requirements on design basis accidents and the use of the TID source tem that is targeted for '
mid-1984 It is also quite relevant to the subject of possible changes in future siting policy. To this end, I am sending a copy of this menorandum and' enclosure to RES for their further use.
0 Odia:554et
!.1 Cze Har 1d R. Denton, Director Office of Nuclear Reactor Regulation
Enclosure:
DISTRIBUTION
- As stated Central Files WPasedag
. ADRS Rdg. JMitchell cc: R. Minogue RS Subject LSoffer R. Bernero RMattson JAMartin HRDenton RBlond DMu11 er ' F& kowski LHulman JMalaro Tn.ny n % ,. trnn. m c p om ge CO'l@
g
.m .
De
DESIGN BASIS RADIOLOGICAL ACCIDENT SOURCE TERMS
.The analysis of postulated accidental events has long been an integral part of the safety evaluations of nuclear power plant designs. These analyses have been instrumental in establishing standards and criteria by which judgments have been made on the acceptability of the functional performance of structures, systems, or components of a facility in response to the initiating and subsequent events. Such events have been .
characterized as design basis accidents (DBA). They represent an
, important part of the design bases for a nuclear power plant as defined in 10 CFR Part 50.2(u). .
In general, DBAs are postulated as credible mechanistic events or event
' sequences derived from extensive engineering and technological
~
experience. Chapter 15 of the NRC staff's Standard Review Plan, NUREG-0800, identifies the DBAs typically. considered in the licensing process for ligh't water reactors. In conjunction with all other design standards and driteria they are frequently thought of as forming a
- design basis envelope w'ithin which the plant is capable of being operated without undue risk to the public. The analysis of accident sequences typically involves a set of assumptions that are explicitly
-intended to bound the physical consequences of a class of similar accident sequerces since it is not possible to analyze all possible sequences. In this manner by consideration of a sufficient number of accident classes there is built up an envelope which is judged to be sufficient to assure the objective of no. undue risk to the public. The assunptions in accident analysis are for the most part conservative, S
.- l-.... ----- -. .. . - - - --. . . . - . -
..o~.
although they also frequently make the analysis highly stylized and unreal, since they 1eflect gaps or uncertainties in the knowledge of some technical. details. Such conservatisms are a part of the Comission's long standing policy of defense in depth.
~
An important'sub-set of DBAs includes those for which a release of radioactive materials within the plant or to the environment are postulated or calculated to occur. For these, the bounding envelope can .
b'e expressed in terms of radiological consequences not expected to be exceeded. Although it has not been the practice to do so, this subset of DBAs could properly be called design basis radiological accidents (DBRA). The, scope of considerations required to be given to accidents postulated to release radioactive material is set out largely in 10 CFR ,
Parts 50 and'100 of the. Commission's regulations. 10 CFR Part 100, "Reaci.or Site Criteria" requires-the determination of exclusion areas, low population zones, and population center distances of sufficient size '
to-limit the radiological consequences to the public of a severe accident. It describes an acceptable methodology for meeting the'se *'
requirements that begins with an assumed fission product release " based upon'a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events, that would result in potential hazards not exceeded by those from any accident considered credible. Such accidents have generally been assumed to
, result in substantial meltdown of the core with subsequent release of l appreciable quantities of fissien oroducts." (10 CFR Part 100.11 footnote) The methodology also specifies that the containment structure 4
i .
i .
-m o. .+ - 4 should be assumed "to leak at its " expected demonstrable leak rate." (10 CFRPart100.11(a))
The deterministic methodology specified by these criteria has led to several descriptive terms for the considered accident. For many years following the publication of the Reactor Site Criteria, it was called a
" maximum credible accident" (MCA) following the usage of TID-14844,
" Calculation of Distance ~ Facto,rs for Power and Test Reactor Sites." .
Some license applicants, adopting the source term assumptions of
, TID-14844, preferred the term " maximum hypothetical accident" (MHA)'. It is of interest to note in this connection that the specification in Part 100 does not. require that the accident " hypothesized for purposes of site analysis," in and of itself, be a " credible" accident. In fact, the accident scenario traditionally specified for site analysis, the so-c'alled TID release, is largely non-mechanistic, not physically possible in some of its detail, and, therefore, not really credible.
The MHA terminology might thus seem 'to be a more accurate description
- for this accide'nt scenario than MCA. '
On the other hand, the Commission's guidance in the referenced footnote to Part 100.11 provides an alternative, i.e., a major accident that can be " postulated from consideration of possible accidental events,-- ."
The TID release scenario reflects a conpromise between hypothetical and pessible events since it was generally the judgment at the time that a necessary precondition of a release of that magnitude would most likely be a loss-of-coolant accident (LOCA). Although the Commission's regulations were subsequently amended to require energency core cooling l
i
) .
systems (ECCS) designed t'o prevent core melting or substantial core degradation, the association of the TID release with a LOCA persists to the present time in Regulatory Guides 1.3 and 1.4 and the staff's Standard Review Plan (SRP), Section 15.6.5, Appendices. Following publication of the General Design Criteria,10 CFR Part 50, Appendix A, the'regul'atory st'aff discontinued usage of the term " maximum credible accident" and the scenario is now often referred to as a design basis accident. This usage is a result of the adoption of the TID' release in the evaluation of the effectiveness of certain engineered safety
, features (ESF) designed to reduce or mitigate the consequences of' fission products released to the containment atmosphere, pursuant to l .
General Desi.gn Criterion 41, " Containment Atmosphere Cleanup." . Credit for such mitigation is generally considered by the staff in computing ,
the release to the atmosphere for purposes of applying the criteria of 10 CFR' Part 100. This coupling of the TID release with a'LOCA and referring to it as a DBA is a source o.f some confusion. A successfully terminated design basis LOCA, i.e., one for which the criteria of 10 CFR 50.46 are not e'xceeded, could not create a source term of the magnitude of the TID release. Use of the latter can be thought of es a surrogate l ..
source tern intended to bound the radioicgical consequences of a LOCA (or other credible accident) that is not terninated before substantial de age to the core has occurred.
i The radiation dose guidelines specified in Part 100 ha've had a sigr.ificant influence upon the deterministic nethods of analysis of the radiclog'ical consequences of accidents. These specify upper linit values to individhals for whole body exposure of 25 P, ens, and 300 Rems 5
, _ _ . _ . - ~ . .__... -- _ - - - -- +- -- --- * " "*-'~---~'
- . -2 0 l
. l to the thyroid from fodine exposure. The potential importance of radioactive foms of iodine es a hazards consideration for the public is thus codified in the Comission's regulations'. These two guidelines were judged to provide suitable bounding values for the potential hazards not expec'ted to be exceeded by those from any accident considered credible, for purposes of reactor siting.
The application of the dose guidelines has typically employe a .
procedure that is consistent with but slightly different from that envisioned in the text of.10 CFR Part 100.11(a) which stated: *
"As an , aid in evaluating a pr'oposed site, an applicant should I
assume a fission product release from the core, the expected
. demonstrable leak rate f' rom the containment and'the meteorological conditions pertinent to his site tolderive an exclusion area, a low population zone, and population center distance."
- s Based on a literal reading, the statement could be construed to requC *'
determination of the (minimum) sizes of exclusion areas and low population zones (site parameters) at which the dose guideline values are just net, proposed site parameters equalling or exceeding these '
values would demonstrate conpliance with the rule. Such a focus would place proper emphasis on the rule as a si-ing consideration.
l L
In practice, an equivalent procedure has been employed because it requires less computational effort. This involves a calculation of dese values for the proposed site parameters and a comparison with the dose .
I
' ~
guideline values. In this case, the proposed site parameters are in compliance with the' rule if the calculated doses do not exceed the guideline values. This practice appears to have had the unintended effect of focusing attention on the resulting dose calculations giving rise to some perceptions that the doses thus calculated and displayed were the " expected" consequences of the initiating events, and their characterization as " acceptable" doses when found to be within the guideline values. Such usage, stands in contrast to the admonition in a .
footnote to Part 100 that cautions that the guideline values of dose were not to be regarded es acceptable limits for emergency doses to any member of the public under accident conditions 6 This practice also influenced t,he emergency planning zone (EPZ) recommendations of.an NRC/ EPA Task Force on Emergency Planning in 1978 (NUREG-0396/ EPA 52011-78-016), subsequently codified in the regulations at 10 CFR 50.47 and.a' revised Appendix E to 10 CFR Part 50.
The TID release postulated to evaluate proposed reactor site paraneters and the credit'to be given to the operation of design features intanded
- to mitigate the consequences of fission product releases is specified in L
Regulatory Guides 1.3 and 1.4 for Bt'Rs and PWRs respectively. The release is characterized as " instantaneous" (from the reactor into the prinary containnent atmosphere) and is comprised of 100% of the core inventory of noble gases and 50t of the iodine. Half of the latter is assuned to " plate out" instantaneously on surfaces inside containnent leaving a net 25% of the iodine core inventory airborne in containnent.
The airborne iodine is assuned to consist of gl% elenental iodine as a a
vapor, 4% as organic iodide as a vapor, and 5% sorbed on aerosols as a particulate form. The original TID release included 1% of the solid fission product inventory as a contributor to direct radiation shine
, through containment. For almost all containment designs, this contribution to dose at the exclusion boundary has been found negligible and it was dropped from the Regulatory Guides. .
Regulatory Guides 1.3 and 1.4.specify the assumptions to be dade .
concerning the transport of ffssion products from the containment to the
, environment. These assumptions neglect or purposely und. restimate most natural processes, other than radioactive decay and leakage, which would reduce the 4,irborne inventory. A1,though the staff's analysis includes atmospheric transport and diffusion, its interpretation includes the assump' tion th'at all transport, of fission products, from the core to the postulated receptor located miles away, occurs instantaneously. The doses to nembers of the public are calculated for two hours and thirty days at the exclusion area boundary and the low population zone,
\ .
respectively. This 30-day period represents the " course of the *.
accident" specified in 10 CFR Part 100. Containment leakage rates are taken at values to be incorporated'in the license Technical Specifications, typically a fraction of a percent per day. Release fractions or source terms into the atmosphere are not commonly recorded for these calculations, the end results usualiy being expressed in terms of radiation dose. As an example, however, for a BWR containment with a leak rate of 1% per day, 0.02% of the core inventory of iodine is calculated as the release to the atmosphere in the first two hours and 0.4% during the entire 30-day course,of the accident. These release
, +
. . t fractions are, on the one hand, substantially smaller than those which have been calculated for severe cere-melt containment failure accidents
~
in PRAs, such "as WASH-1400, but are substantially larger than one would calculate for a large-break LOCA with delayed emergency core cooling system operation, such as the AD-1/2 sequence described in NUREG-0772.
As noted above, the containment system as an engineered safety feature ;
(ESF) is always credited with a substantial consequence mitijation effectiveness. It should be n'oted, however, that the specified
, acceptable leakage rates are not independently derived quantities, but rather are values judged to be con'servatively low and reasonably ,
[
achievable w,ith due care in design, and maintenance. One result.of the t
calculational procedure and TID source term assumptions is that the .
! iodine, (thyroid) dose has typically more closely approached the Part 100 '
guideline value of 300 Rem for the first two hours, and usually for the i courseofaccident(30-day)doseaswell. A substantial reduction in '
the iodine source tem, which may well be more realistic, would have the
! N l effect of decreasing the acceptable size of exclusion areas, or * ' '
^
increasing acceptable containment leak rates (or both) and would be ^
1 . :
I governed solely by the noble gas releases, assuning no other changes in regulations or staff practices were involved, t ,
There are other ESFs that can be involved in the analyses used to I establish compliance with siting criteria. Atnospheric cleanup systems within containnent have typically been incorporated in PWP designs, i inese are modificatiens of systems designed to remove heat from the -
containr.ent building and include sprays, ice condensers, and, in a few e
g g
--+, .._-_______._._--___n_.._..
. cases, recirculating filter systems. Staff practice has generally required that PWR containment sprays and ice condenser systems contain chemical additives to enhance the alkalinity of the spray water or ice-melt to increase the effectiveness for absorption of elemental
. iodine. These systems are not considered to be particularly effective for removal of organic iodides and moderately effective for the removal of aerosols. Credit given for the cleanup capability of such systems is interactive with containment l'eak rate and proposed siting pa'rameters ,
and typically also leads to sp'cifications on certain aspects of system e
, design.(e.g., both the design and locations of spray nozzles) and -
performance (e.g., automatic switchover from injection to recirculation -
of containment sprays). A substantial change in the TID release with respect to the chemical character of fission product iodine in '
con, tai.nment w'ould require a reanalysis of an appropriate design basis for these systems. '
Pressure suppreysion pools in BWRs are designed to condense the steam released to the drywell (primary containment) during a LOCA. In
- contrast to the case with PWR containment heat removal systems, they ,
have not been nodified to enhance fission product removal and the staff '
has not credited BWR suppression pools for containment cleanup in its analyses directed to compliance with siting criteria. Realistically, hcwever, one must expect that the transport of fission product contaninants with the air-steam mixture entering a suppression pool would result in a scrubbing effe:t and some retention in the suppression pool water. If quantitative credit were to be given in the future for t
fission product renoval by BWR s' p;ression pools, an appropriate design f
e
~
, basis source term would have to be developed. There are, however, other ESF systems for BWRs that are currently credited for mitigation of '
ace'ident consequences. One of these is a main steam line isolation valve leakage control system (MSIVLCS). The main steam lines in a BWR represent a potential leakage path from the reactor pressure vessel directly to plant structures outside both primary and secondary containment barriers. Multiple isolation valves are provided to seal these steam lines and leakage control systems are provided in .
contemporary plants to control the leakage through these valves. The
, current regulatory requirement (Regulatory Guide 1.96) for MSIV LCS is traceable to the credit given for mitigation of. iodine exposure inherent .
in the TID re, lease. ... ,
Most leakage collection systems operate by drawing a partial vacuum ,
in the, volumes.between closed isolation valves and exhausting through a e filter train. This is usually accomplished by connecting the exhaust to the standby gas treatment system (SGTS) which contains filters. The effectiveness of MSIV leakage control systems depends upon the maintenance of very low MSIV leak rates and upon the effectiveness of **
the SG,TS. The SGTS is a ventilation control system that traps contaninants leaked from the primary containment and collected in the secondary containment or reactor building. The filter train typically
. includes noisture separators, heater, pre-filter, HEPA filter and a charcoal trap. The charcoal trap is the dominant feature credited for elemental iodine adsorption. In this, as in other ESF filtration systems that have been required, e.g., in centrol roen, enclosure building, and spent fuel storage building ventilation systens, the i
relative importance of the particulate (HEPA) and iodine (charcoal) i .
l
~
? .
.. filters may be altered with changes in the design basis (TID) source te rin. I The licensing process also involves analyses of a number of postulated
~ accidents with potential for fission product release that serve other safety purposes related to plant design and performance and onsite radiation protection of workers. The regulatory basis for these is found largely in 10 CFR Part 5'0, including several of the Ge eral Design .,
Criteria (GDC) of Appendix A to Part 50, and in post-THI requirements set out in NUREG-0737. Particularly relevant design criteria in '
addition to GDC 41 previously mentioned include the following:
GDC 4 " Environmental and missile design bases." Postulated
,. accidents are used to establish an appropriate basis for defining the, radiation fields in which equipme'nt important to safety must be capable of functioning.*
l \
- The new rule on Environmental Qualification of Electric Equipment (10 CFR 50.49) nore explicitly identifies the need to consider radiation t
environments including those fron design basis events.
l I .
1 1
l
GDC 19 " Control Room." Sets a radiation protection design criterion for access and occupancy of the control room under postulated 3
accident conditions.
" Control of releases of radioactive material to the environ-ment." Requires consideration of releases from an icipated .,
operational occorrences, i.e., events which are expected to "
, occur one or more times during the life of a nuclear power unit. - -
GDC 61 " Fuel storage and handling and radioactivity control." Re-quires consideration of postulated accidents in the design of fuel storage and handling systems.
GDC 64 " Monitoring radioactivity releases." Requires monitoring of plant areas, effluent paths, and the environment for
- radioactivity releases from normal operations, anticipated operational occurrences, and from postulated accidents.
The postulated accidents and accident source terms used to evaluate compliance with these design criteria, described in regulatory guides and the Standard Review Plan, NUREG-0300, are discussed below. In almost all cases.. the dcminant hazards consideration h65 been associated with radioicdine and the treatments strongly ref1r.ct the influence that
r
- ~
~
. . :. .ip
/ -
the site evaluation criteria and the TID release assumptions have had.
As in the case of the application to site criteria' discussed above,
~
these analyses have been deterministic in nature,Iave employed conser'vative assump[fon:,, and thrresults typically expressed in terms of dose. It must also be noted that some of these analysis practices have also led to specific license conditions in the forrn of technical specifiiations pursuant to 10,CFR 9 art 50.36. 2
~ ,
. In the post-Three Mile island era, some additional licensing
~
, requirements ' involving accident source terms have been identified it NUREG-0737 and some are codified in the CP rule at 10 CFk 50.34(f). The primary. thru.st of these has been directed toward improved radiation prot'ection of'onsite personnel, stricter standards for the qualification of equipment in a post-accident environment, and improved capability for instrumentation to follow the course of an accident. A~sunmary listing of licensing documents involving the postulated TID release or variants
- thereof is givEn in Table 1.
There are also other fission product release " scenarios" for otherj postulatedaccidentswithinthe'designbasfIthathavebeenusedto establish criteria for a variety of safety purposes. The Standard Revie ' Plan sections in which these are addressed'a e '.isted in Table 2.
in general, these latter postulated accident: care cons'idered as reasonably likely to occur. The source of fission product a:tivity is typically that which could be present in the circulating primary coolant or, in some cases, that which could be released frcm fuel rods if clad I
rupture is predicted.to occur. Although not stecific.hlly provided for I
8 I
(
, in the regulations, the staff has generally applied dose criteria in terms of fractions of the guideline" values of 10 CFR Part 100 for these accidents. Th'is practice, coupled with the fact that radiciodines are normally present in reactor coolant and are known to be released if clad ruptures occur, also results in a dominant role for thyroid dose calculations. The chemical form of the iodine has not, for the most part, been a significant factor in the calculational procedures in these applications. -
~
, Common to all of the treatments of postulated design basis radiological accidents are the principal elements of the consequence modeling. These include atmo, spheric transport and , dispersion in which the entire release is treated as a gas, non-condensing vapor, or extremely small particle size aeroso1'such that n'o ground deposition is assumed to occur.
, , The
~
resulting dose calculations reflect only direct radiation exposure from the airborne cloud, and the inhalation of the airborne radiciodine that '
leads to thyroid exposure. Other potential pathways of radiation N
exposure to individuals are not treated, nor has it been the practice in
- safety evaluations to calculate population exposures (man-rem). This stands in marked contrast to the consequence modeling currently enployed in probabilistic risk assessments in which more pathways are considered and the radiation exposures involved are expressed in terms of potential health effects.
~
9 4
e 4
9 e
_____m_m.__.- -_
, , , , , ~ *
- Table 1 Licensing Documents Involving a TID Release Assumption Part 1 - Enoineered Safety Feature Perfomance '
Subject . Regulatory Standard Review Guide Plan Section Containment Leaka'ge 1.3, 1.4 6.2.1,6.2.6,6.5.3 Containment Sprays 1.3, 1.4 6.5.2.16.6.5 App. A Ice Condenser ---
6.5.2,6.5.3,6.5.4 Pressure Suppression Pool ---
6.5.3 P.5IV Leakage Control 1.3, 1.9.6 6.7,15.6.5D SGTS 1.52 9.4.5,15.6.5 App. D Aux. Building Filters ,
1.52 6.5.1,9.4.2,3,4
- Containment Recire. Filters 1.3.1.4.1.52 6.5.1,15.6.5 App. A Control Room Habitability Systems ---
6.4
~ .
(Also 10 CFR 50.34(f)(1)(xxviii) and NUREG-0737, Item III.D.3.4) l
\
Part 2 - Accident and Post-Accident Environment '
~
Equipment Qualification 1.89 3.11 (Also NUREG-0588)
Combustible Gas Generation 1.7 6.12,6.2.5 Plant Shielding --- ---
l (10CFR*50.34(f)(1)'(vii),andNUREG-0737.ItemII.B.2)
) Post-Accident (10CFR50.34(f)(1
- Sampling)(viii)andNUREG-0737 Leakage Outside Containment .ItemII.B.3) 15.6.5 Apo. B (10CFR50.34(f)(1)(xxvi)andNUREG-0737,ItemIII.D.I.1) -
N i
/
^~'
y ,
Table 2 - Postulated Design Basis Radiological Accidents Identified in Standard Review Plan (non-TID Release Assumptions)
SRP Section Sub.iect 15.1.5 App. A Main Steam Line Failure (PWR) 15.2.8 Feedwater System Pipe Breaks (PWR) 15.3.3/4 RCP Rotor Seizure or Shaft Break 15.4.8 Rod Ejection Accidents (PWR) 15.4.9 Rod Drop Accidents (BWR) 4 15.6.2 Failure of Small Lines Carrying Primary Coolant Outside Containment 15.6.3 Steam Generator Tube Failures (PWR) .
15.6.4 , MainSteamLineFailure(BWR) 15.7.3 Liquid Waste System Failures 15.7.4 Fuel Handling Accidents
. . 15.7.5 -
Spent Fuel Cask Drop Accidents 9 '
e e
\
4 e
l e
>w