ML20149F799
| ML20149F799 | |
| Person / Time | |
|---|---|
| Issue date: | 06/17/1985 |
| From: | Hulman L Office of Nuclear Reactor Regulation |
| To: | Bernero R Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20149B718 | List: |
| References | |
| FOIA-87-743, RTR-NUREG-1150 NUDOCS 8802170369 | |
| Download: ML20149F799 (59) | |
Text
~
.' ~ ? ,
- y . .z'o, UNITED STATES i ! NUCLEAR REGULATORY COMMISSION F;' ! ( $ WASHINGTON, D. C. 20665 y;- \,**..e/
S. , , , 17 M DEI
/1) NOTE T0: Robert M. Bernero, Director L,'y . Division of Systems Integration EI) THRU: Daniel R. Muller, Assistant Directogd /
- f. for Radiation Protection (y ~ ly' b Division of Systems Integration V i'l 1
[
FROM: L. G. Hulman, Chief Accident Evaluation Branch
. Division of Systems Integration
SUBJECT:
STATUS OF PROPOSED COMMISSION PAPER ON REGULATORY VSES OF
'I',
SOURCE TERM RESEARCH s
Enclosed is a copy of Draft 3 of the proposed paper. Still to be integrated into the paper are the source term interrelationships with the RES risk rebaselining and the IDCOR/NRC activities to be supplied by Rosztoczy. Coments on Draft 2 were solicited from a wide sector of management. Incorporated in Draft 3 is my resolution of coninents from the following individuals who responded: .
D. Eisenhut J. Malaro G. Arlotto W. P. Gancill E. Jorcan J. Mitchell G. W. Kerr S. Acharya C. Xelber T. Quay D. Muller L. Soffer F. Rowsome J. Read
.. Four major comments that I did not resolve in favor of commentors, and the reasons why I did not, are as follows:
Commeg (Malaro) the proposed changes in accident assessments and their use are so pervasive that significant changes in f regulations and regulatory guides are required before implementation can be effected.
x N
Q/C1 pgg21g9geopo4 SHOLLYB7-743 PDR #
____J
- ' ,e ,
- .s ,
0 gA k:
O ' Response M M While the proposed changes may be pervasive, hey do not constitute a radical departure from present practice except
?>.J for proposed rulemaking related to siting and emergency if .-
planning. To the contrary, the staff has made a significant number of changes in practice in many related areas without
. changing regulatory guides before implementation, or calling i for rulemaking. Furthermore, 10 CFR 100 provides for adjustments in related practice in a footnote which states in 6 part
<[ "The calculations described in Technical Information Document 14844;may be used as a point of departure..."
- f. f.
,,' Coment (Rowsome) The use of a deterministic approach to reactor licensing should be abandoned in favor of a maturing risk based methodology
Response
For new plants and designs I agree the approach may have merit.
Indeed, some of the forward looking activities in NRR and RES can provide bases for revisions to practice in this area.
For the existing suite of plants, however, I do not agree. The primary reason is that our practice constitutes a reasonably predictable basis for licensing new plants and regulating operating reactors. A major change in practice of the type
. suggested, in try opinion, could significantly destabilize the licensing process for the existing suite of operating reactors
. and those scheduled to operate in the next few years.
~
I have considered. the above approach in the discussion of new regulatory requirements as an alternative to the present design basis accident methodology.
Comment (Rowsome) The existing emergency planning regulation and practice
" should be abandoned, possibly in favor of only pre-planned
. f licensee and Federal response (with ad hoc responses from state and local jurisdictions). The three primary bases appear to be that historical ad hoc evacuations appear to have worked adequately,
g .,
~
A. . ,
J' ' . .. .
j m.
c.t t : C.
u:
ki that risk-based cost / benefit assessments of regulatory activities R in the subject area were unlikely to be cost beneficai, and that sheltering in lieu of evacuation appears to be a bt.tter
- 4) 4 strategy.
G.
Response
.id':
i I conclude it is prudent to plan to fine tune the practice in the j:: subject area, but not to radically oveihaul it along the lines suggested. The reasons are that the emergency planning area j{
t suffers more from an uniformed public than it does from regulatory fi guidance and execution; that alternatives to evacuation may very
- 1. .' well be better than evacuation in many but not all situations; and
,,. that it is too early in contemporary risk rebaselining to conclude
,. that across-the-board changes in planning guidance and execution for every reactor site are cost beneficial or not.
? Coment (by a number) A wide range of views on the adequacy of research
- to support decisionmaking were offered. The views ranged from sufficient bases have existed for years, to wait until all the
.research is completed.
. Response The proposed paper is intended to be a staff plan for the integration of research findings into licensing activities over c the next three years as they become available. The adequacy of those findings to support decisionmaking is and will continue to .
be controversial. However, I conclude that licensing decision-making must move fomard since several findings to date appear sound, others appear to be maturing and sound planning must be fomard looking.
With the exception of integrating the long-awaited Rosztoczy input,
. I consider the enclosed draft ready for senior management review and presentation to the Comission. Some of the concepts in the draft, duly qualified as not being staff recomendations or i positions, are proposed herein for presentation to the ACRS on August 2. I understand that many of the suggestions for implementation contained in the draft may be considered not f supportable without the results of NUREG-1150 in hand. Furthermore,
,4 e
.s, ,,
1 f.' . p.4
'O y .,
,. I understand regulatory implementation is planned to be a
[.f; .0 primary conclusion of NUREG-1150. If many of the implementatfor.
7 ;;* activities discussed in the draft Comission paper must await
. .: . NUREG-1150, many benefits may be lost or unnecessarily postponed.
'1 In particular, backfitting iodine filters, automatic containment Wi spray additive injection and considerations of containment leak rates
'j
l and integrity appear to be worthwhile areas that can move forward.
f.'g -
I await your instructions for further action.
. Dm lJ. L. G. Hulman, Chief Accident Evaluation Branch
.Y. Division of Systems Integration 5:
Enclosed
- 1) Draft 2 of Proposed Source Term Commission Paper cc: w/ encl.
' T. Speis H. Thompson
,. D. Muller E. Jordan J. Read A. Spano -
F. Eltawila
. J. Malaro F. Pagano/E. Podolak J. Rosenthal
, J. Mitchell
. R. LaGrange J. Saltzman 9
4 e' - ) _
i e
q .
1 l
i l
- - = - .
t '.
O *#
,.s, i .
!.l .
f ,
h:::
If(! -
@J For: The Comissioners ki?
b From: William J. Dircks Executive Director for Operations
. 3-
.; .}-
Subject:
STAFF PLANS FOR THE USE OF NEW SOURCE TERM INFORMATION IN THE
{l REGULATORYPROCESS(DRAFT 3) ,
jl
Purpose:
To inform the Comission of current and forthcoming staff
- activities for use of new source term infonnation in the regulatory process; to infonn the Comission of preparations for Comission actions based upon new source term information, and to enable the Comission to coment on the direction and p'riorities of these staff activities.
~
Sumary: ,
In another document, the staff has completed NUREG-0956, Reassessment of the Technical Basis For Estimating Source Terms.
That document sumarizes and assesses recent advances in what is called source term research; that is, our ability to predict the quantities of radioactive materials transported and released i'
Contact:
R. M. Bernero, NRR, 492-7373 or L. G. Hulman, NRR, 492-7763 4
E ....
~ .. ,
i e'
m
.. a '
q', '
(O in reactor accider.ts. The ability to predict, or to assure t.
(
?? conservative assessments of these releases lies at the very .
Vr
'(j heart of the regulatory process. We measure the need for and
... , . .j ', , 13 risk outliers. Furthermore, many remaining experimental, code ;4,' m, develepment, code verification and uncertainty research activi-1), ties are directly related to the ability of NRR and IE to e implement research findings. I conclude that the framework for "..c. . the orderly implementation of source term and related research .. into the licensing process described herein provides for the necessary timely interactions between the program offices, and between the staff and the Comission. Finally, the staff offers +.he agenda in this paper for whatever coment or direction the Comission may wish= to offer. Scheduling: This information paper should be presented to the Comission at the same time that the ASTP0 sponsored research culminating in NUREG-0956 is e.plained. An open session is recomended. William J. Dircks Executive Director for Operations
Enclosures:
- 1. IDCOR/ Staff Interactions Relating to Source Tenns
- 2. Nine Other Regulatory Areas and Potential Changes
- 3. Preliminary Cost / Benefit and Schedule for Change 4 Draft Bra ,ch Technical Position on Control Room Backfits
- 5. Draft Branch Technical Position on Automatic Injection of Spray Additives for PWRs
.,y: .,
', .- . ... . r m e
t
?;
REGULATORY AREAS AND POTENTIAL CHANGES
-).
I- A review of existing licensing or licensing related areas of regulatory J.' practice influenced by accident source term assumptions was used to identify areas where the staff concludes changes may be appropriate. Ten areas have
. been identified. Enclosure 1 indicates the use of source term information in the search for ris'k outliers. For each of the other nine areas identified, below, a description of the background and current practice, possibility for and character of potential changes in practice, costs, benefits and schedule follows:
- 1. CONTAINMENT PERFORMANCE Background and Current Practice: The current design basis leak rate of containments ranges from 0.1% to 2.0% by volume per day. Licensees are
~
required by 10 CFR 50.54(0), to perfonn an integrated leak rate test in accordance with 10 CFR 50, Appendix J. These tests are performed at least once every forty months to demonstrate that the actual leak rate is less than 75% of the design basis value contained in each plant's technical specifications.
- In addition, the gas and liquid leakage from equipment and components that may handle post-accident contamination outside the containment is addressed I
(TMiActionItemIII.D.1)byaleakdetectionprogrnm. l l l .
E], . .
.,. ._ , , .. -3 ,
-; )
. [.
r-j i 7,1 The design basis leak rate of a containment is established by calculations
.?
- . ' which assume that the post accident airborne concentrations of iodine vapor y and noble gases specified by TID 14844 are instantaneously released q
and uniformly distributed in the containment atmosphere. These calculations ('; deliberately ignore or treat simplistically the attenuation of iodine by such natural processes as dissolution in water, chemical reactions, and plate-out (adsorption). They also neglect release of other fission products such as 6 cesium, tellurium and strontium, and non-radioactive aerosols to the
- containment, and any undetected breach of containment.
In progress at this writing are proposed changes to Appendix J that do not incorporate new source term information. Possibility for and Character of Changes: The assumed release of iodine vapor during reactor accidents currently dominates the calculations of off-site doses from which technical specifications of containment leak rates are derived. Current research indicates that most iodine released in core-melt , accidents will be in chemical and physical forms other than iodine vapor, and that elemental iodine does not dominate off-site risk. It is also now known that aerosols of many other elements released following core melt accidents, and containment by-pass or failure, not diffusive leakage, are of greater t importance to public risk. Thus, the basis for specifying containment leak rates should be re-evaluated, and a greater emphasis should be placed on , maintaining the integrity of containments. 9 l
?4 e *
.s p,t - 3 a -
~'
In the near-tem, the_ staff intends to study the advantages and e.;
.5 disadvantages of alternative containment leak rate test requirements and acceptance criteria in terms of accident risk significance as well as X operational considerations. Releases of noble gases, essentially unchanged by present source term research, will be used to establish an upper limit for containment leak rates since releases of other isotopes appear unlikely to
[ dominate offsite doses. Early indications are that these leak rates might berelaxedbyfactorsrangingfromtwototen, depend $nguponthepresentvalue, without significantly impacting public health and safety. Also to be considered are management techniques that would delay or prevent containment failure during severe accidents. On the basis of this study, the staff may propose revisions in the Standard Review Plan (SRP) criteria and in the Appendix J test requirements and acceptance criteria. Candidate revisions may include:
- 1. Specifying a containnwnt leak' rate value and testing freq.sency for each of the major types of containment.
~2. Adopting administrative controls or penetration design guidelines to aid in precluding an undetected breach of containment integrity.
- 3. Requiring testing to provide additional assurance that an undetected breach of containment integrity does not exist (inherent design / operation features such as the need to maintain a subatmospheric condition or
lb 3:/'
.N .inerted atmosphere may be readily adapted for this purpose). Licensees h will then have the option of utilizing the new criteria in technical specification changes.
- 4. Modify staff criteria for closure times for valves in lines which pene-
,' . trate containment and may be open during operation.
e ,~
,- To assess the extent of any proposed revisions, the following actions will be '. considered:
- 1. Perfonn mechanistic analyses'(releases from fuel, plate-out, spray washout, attenuation along release path, reemission etc.) of fission product releases from the core and transport within the containment.
- 2. Determine threshold levels of containment leakage'important in risk estimates.
- 3. Explore the viability of implementing testing practices that provide a continuous indication of containment integrity.
- 4. Provide bases for adjusting Limiting Conditions for Operation and ActionStatements,intheTechnicalSpecifications(i.e.,toreflect theirimportancetosafety).
1 I
- +c-3-~. . . . . . . . ,
?H 5 bj For the longer term, the staff proposes to:
- M n .!
t
@ 1. Explore the use of containment venting schones for PWR's (similar to that - proposed for BWR's) to reduce public risk by preserving containment ' 'g) .. integrity.
o
- 2. Examine the desirability of certain surveillance / maintenance practices
~
(such as perfoming local leak rate tests on containment isolation valves during operation modes 1 to 4) that might lead to inadvertent post u accidentsafetysystemlockouts(suchascontainmentsprays),anddevelop an operability check method for preventing inadvertent safety system lockouts. . Costs and Benefits of Change: Near-term changes exploring the risk significance of the alternative leak rate criteria in light of recent source term research are expected to require moderate staff resources. Longer term efforts, especially those investigating approaches to enhanced containment 1 L integrity may require a greater involvement of staff resources. I Increases in allowable containment leak rate of the order of factors
~
- - ' of two to ten are expected to offer direct and significant regulatory p! relief to many operating licensees, since compliance with the criteria L I of Appendix J is costly and is incurred regularly. The expense involved in the testing becomes arger as the leak rate to be l l8 i i l ( .
4 , 1 ,
.; 6 n' '.
demonstrated becomes smaller. Furthennore, changes in monitoring
.":. criteria can be upected to reduce the frequency of integrated tests. ;T[;
T, Imposition of additional containment integrity requirements would have
' costs ranging from relatively low for routine administrative controls ']
to high for systems which continuously indicate the status of containment
- integrity.
i , T- Tenative Agenda and Schedule: The staff's near-term study and recommendations are expected to begin in late FY85 and be completed in FY86. Longer. term studies will comence in FY86 and are expected to be completed in FYB8. 9 4 6 e i., 8 i i
- 1, . m,
;, . l -
. _ . . .c , ,
a 4 L5 N .
]: ; 2. ENv1RONMENTAL QUALIFICATION OF EQUIPMENT:
rf Background and Current Practice: 10 CFR Part 50.49 requires electrical 7 equipment important to safety to be capable of remaining functional during and following design basis accidents for which that equipment is needed. The
'r: radiation doses assessed for this equipment in an accident environment are derived from the TID-14844 assumptions. In addition to the source term . }- 4 I research is a great w'ealth of information emerging from the examination of I equipment subjected to the accident environment at Three Mile Island Unit 2 that can be useful in assessing operability.
Poss'ib111 ties for and Character of Change: The radiation dose rates and integrated dose values arising from new source tem research are likely to , differ from those derived from TID-14844. The new information may show that - equipment would be exposed to lesser quantities of iodine than previously postulated, but exposed to other fission products presently neglected. At the present time it. is not clear whether the source term research will lead to an increase or a decrease in the accident radiation, temperature, pressure, moisture and particulate environment for equipment qualification purposes. The staff estimates, however, that the present degree of equipment qualification, derived from TID-14844, provides a substantial level of protection for many severe accident conditions. A study wil'1 be perfomed to compare the margins , provided by the existing equipment qualification criteria with those derived from new source term data. The information emerging from examination of 4
- the Three Mile Island Unit 2 equipment will also be considered. If the study L i 4 - - - , - - - ,
MT .. q,v . , . q; - 1; .. y* , ., , n :. h 8
- 5.
- 93:
7:,1, indicates that the new source term data would result in a higher level of
.~ 1
- 9. radiation for somt accident sequences.. then the study will also addrsss the 4
risk significance of having equipment qualified only to the lower level.
- [,.~ .%
"., Costs and Benefits of Changes: It is difficult to provide a clear statement ,
i
," of costs and benefits since it is not certain what changes may occur.
s i. The staff estimates a moderate to high effort to be required to evaluate
/,' the margins associated with the present criteria, and to address their risk significance with respect to the new source term research. .
If the present criteria prove to be conservative, the staff antic,1 pates no direct relief to the industry, since there would be no great impetus to replace presently qualified equipment. However, at future maintenance outages, equipment with superior performance characteristics, but only capable of being qualified to less harsh radiation environments, could be used to replace present equipment. Also, in some cases, it may be possible to extend the qualified lifetime of equipment. It is unlikely t that qualification to moderately lower radiation levels would significantly ! reduce costs. If the present criteria are non-conservative in radiation level, but have j 7-little risk significance associated with earlier equipment failure, the staff does not anticipate requiring licensees to take any imediate ! corrective action, such as replacement of equipment or its radiation l i
7:. ;
,a 9
h., s Y:! sensitive components, but might require equipment or components of (({ hb! equipment qualified to highe' radiation levels 'o be installed gradually, en )j x. during periodic maintenance, as older equipment or components are replaced.
- 37. .
This could potentially represent a large cost to the industry, depending on .;c how non-conservative current criteria may be, the number and type of 5 components involved, and how much further testing would be required. YI s If the risks from exc'eeding the present criteria (radiation level temperature.
- pressure and aerosol loading) are judged significant, the expected impact on the industry would be high since prompt corrective action and/or shutdown might be required. The impact of any increase in radiation levels above current criteria would have to be assessed on an equipment specific basis; i.e.,
. different items of equiprent are currently shown to be qualified for ,
different levels of radiation. i Tentative Agenda and Schedule: A staff study to address the margins provided 'l by the existing equipment qualification criteria will be initiated sometime in FYB6, and completed in FYB8. Although the staff can assess the change in criteria on a generic basis, involvement of industry may be required to assess the impact of a criteria change on specific items of equipment. It should be noted that changes in qualification to incorporate magnitude or
- duration of temperatures / pressures associated with severe accidents could !
haveasubstantialimpactonquaiification. 1
--- ,, , . . . y.
. - - . .n . .- . .. .
10 E lh 3. . EMERGENCY PLANNING
;e ,;
Is..; Background and Current Practice: Current emergency planning regulatory y;: M practice is based upon consideration of a spectrum of accidents ranging from
'sJ
', ' : relatively frequent but low consequence events (including design basis accidents),tolowprobabilitysevereaccidents. The basic document
,, upon which the planning requirements are based. NUREG-0396, evaluated severe E', - accident consequences'and probabilities using WASH-1400 severe accident source u.
' '. terms as the fundamental assumptions. Subsequently, a regulation setting forth a generic approach to onsite and offsite planning and preparedness (10 CFR 50.47 and Appendix E thereto) was promulgated. Generic requirements
.were set forth for the establishment of a plume exposure emergency p lann ing .
t zone (EPZ)ofabout10milesinradius,andaningestionexposureEPZofabout 50 miles. The sizes of these zones were considered sufficient for the planning of various possible protective actions at any given nuclear power plant.
- Acompanionguidancecriteriadocument(NUREG-0654),preparedjointlybythe NRC and FEMA, promulgated guidelines that included botti evacuation and sheltering as potential offsite emergency responses. Since the NRC emergency planning requirements include actions by state and local governments, the NRC depends upon FEMA to assist in providing an evaluation of the state and local offsite plans and preparedness around each reactor site.
Possibility for and Character of Change: As a result of the generic approach f to EP, misconceptions have arisen that all persons living within the EPZ are at high risk and that those living beyond the 10 mile limit are safe from I h
- - ~.- .; . _ a , ~ w
'"w .c~
8
% e
,;..' 11 :
, = *;
I k.il :
~,
g radiation exposure. In addition, some have interpreted the regulation h . as virtually always requiring evacuation to 10 mi.es or even further. The , I
+d use of WASH-1400 source terms to study emergency planning indicated that ;
i
@4 the potential for early injuries and fatalities from severe accidents has f a large variation within the EPZ; that is, the risks are much greater in the "c inner portion close to the reactor than at the perimeter of the 10-mile EPZ. .
- The concept of a graded response that would recognize such a risk variation
+
- r. i
- t. with distance and time was beginning to gain support at about the time the
'4 peer review process for the revised source tem research began. Information ,
obtained from the new source term research clearly supports and confirms this ; concept. Further1nore, the results of the source term work indicate that there ' may be a reduction in offsite accident impacts that may be specific to plant and site characteristics. Thus, employing generalizations or generic requirements for all plants may not be technically valid. The source tem research also indicates,that severe accident releases are more time dependent [ than those calculated using the WASH-1400 methodology, a conclusion which , affects the timing of any needed emergency actions. Proposed Actions: A sequence of activities is proposed to implement changes in requirements and practice as follows: l
- 1. Coordinate (IE) with FEMA and solicit their assistance in meetings with ,
appropriate state and local officials to explain the "graded response" t approach to emergency planning and the possibility of classifying I plants by groups by rtlated accident consequence and risk estimates.
. i 1 ,
E : l .i . l
)
.. - - - - ~ . ,, . . , - , - -l?. , , - -- . - - - .
- e ,
4
*) 12 m
Eb h 2. With FEMA's assistance, detemine (IE) the impact of a proposed change to emergency planning as discussed above, including apt, opriate protective action response strategies. N). } :*1 g :! 3. Draft (RES and IE) revisions to 10 CFR 50.47 and Appendix E, taking into [ consideration the revised source tem methodology and using the following alternative approaches: (1) classify plants or groups of plants according e
'[
to their apparen't risk profiles., (2.) use site specific emergency plans, or (3) use the graded response concept.
- 4. Usingaconstituencydevelopedonanalternative(IE,RESandFEMA)begin a parallel effort to the rulemaking to revise guidance contained in the joint FEMA /NRC NUREG-0654 document.
- 5. Provide (NRRandRES)detailedtechnicalanalysesandrationaleforchange (NUREG-1082) a's a result of the new source term infomation. This document will fem the cornerstone for EP changes.
L
- 6. Coordinate (IE) with EPA to ensure that any NRC actions do not conflict
[ with the EPA ongoing effort to revise the Protective Action Guides. Costs and Benefits of Change: Changes to emergency planning are expected to require a moderate to high effort in staff resources. Significant elements in this effort will include (1) development of the technical bases for change (NUREG-1082) resulting from revised source tems and risk profiles and
)
4 1
- u. ,
Is; 13 y,f !(j ,'. (2) coordination with other federal agencies, such as FEMA and EPA, and with affected state and local agencies. Rulemaking activities, specif 'cally with , ((:j an anticipated hearing, will require the moderate expenditure of staff .. : e t "' resources. >:S n f_ Tangible benefits in terms of direct regulatory relief to operating plant C* licensees are expected to be low, since the staff anticipates no major changes in licensees emergency response capabilities or facilities.
-
- Intangible benefits in terms of reduced risk perceptions are anticipated to be high. Benefits to state and local agencies involved in energency planning and response are expected to be high, since it is anticipated that a significant portion of the planning efforts devoted to the peripheral region of the EPZ may be eased. Revised emergency planning criteria more closely linked to our best understanding of accident risk can be expected to significantly enhance public confidence and lead to a more stable and efficient licensing procedure, as well.
Tentative Agenda and Schedule: The technical basis for change is expected to be initiated in FY 1986 and rulemaking could be completed in late FY 1987. 4 e f e t G G
= m ~
. c. . . s_.s .
3,- 14 N
.i 4. ACCIDENT CONSEQUENCES AND INDEMNIFICATION
[?. {.f.
- h. 3 Background and Current Practice: Nuclear power plant licensees are required
- E
- ,
by 10 CFR 50.54(w) to obtain the maximum on-site property damage insurance reasonably available, and by the Price-Anderson Act of 1957, as amerided, e are required to obtain the maximum liability insurance coverage available, e. .',. Should an accident at any U.S. plant incur offsite liabilities in excess of
~ . ! '? that coverage, a retr'ospective premium would be paid by all licensees to 3.. ,
I create a fund to discharge this excess. At present, these two layers of
'~
offsite insurance provide $630 million of coverage, an amount that increases' by $5 million as each new nuclear power plant is licensed. The effective liability insurance limit and the retrospective premiums are set by. Congress, and do not directly reflect an actuarial value of t' eh indemnification provided. The Comission does, however, periodically provide recomendations and l information concerning accident consequences, probabilities and liability limits to Congress. An example of such information is NUREG-0957. Noted f
, therein is the conclusion that the two layers of insurance should' provide sufficient liability protection for most accidents, but there remains a very -
low probability of high consequence events that could result in public liability claims well in excess.of the available insurance. Also, whether i, source tenn research will change this risk conclusion has not been ,
~-
established. In the event of an accident, the Comission must make a detennination of (
, whether or not that accident was an "extraordinary nuclear occurrence" l
1
"je.' 15 Q-using criteria contained in 10 CFR 140. The radiological criteria in 10 CFR b.,1 4
- f;' 140 correspond to,much lower severity levels than the dose guidelines of 10 CFR
- n 100, and could, in theory, be exceeded by accidents not involving core damage.
.L? The accident at THI-2 was, by 10 CFR 140 criteria, not an extraordinary nuclear 3.,
'( occurrence. The Comission is considering revisions to the definition of an "extraordinary nuclear occurrence." 4- .' s T Off-site risks of both human injuries and property damage are estimated by the staff for each plant for purposes of implementing the National Environ-mental Policy Act, and are reported in Draft and Final Environmental Statements prior to licensing. - Possibility for and Character of Changes: Amendments to the on-site property insurance provisions of 10 CFR Part 50 are being considered. To detemine what the impact of the new research would have on regulatory requirements for such insurance, two tasks would be undertaken. First, bases for the existing rule (NUREG/CR-2601) would be assessed in light of the new infomation. Secondly, depending on the outcome of the first step, modifications to the existing rule [ would be proposed. The current Price-Anderson Act expires on August 1, 1987.
- The Comission has recomended that Congress extend the act and amend it to remove the liability limit. This last would be done by providing for retro-
[ I spective premiums to be paid every year following an accident until all claims are settled, and by doubling the premium to $10 million per operating power O reactor per year. It is possible that the future availability of the results of a thorough re-examination of the risks of severe accidents, using the new 4
E, , v .
- p. d 1.l 16
?.j .
fi, iI methodology, may affect the outcome of present or future Congressional
.Y!
f,'; deliberations of this matter; perhaps suggesting different treatment of certain
.p
',- reactor / containment types, or a decreased need for or a reduced cost of
- indemnification.
. t ., .
2:
;,5 The improved source terms will be used to compute more realistic off-site 'T[ consequences in future NEPA reviews.
7 - [ Benefits and Costs of Changes: Impacts on staff resources are expected to be minimal. Since the risks of nuclear power are small, the expected costs and benefits to individual utilities governed by the Price-Anderson Act are also small, actuarial?- In the event of an accident, however, the difference between limited and unlimited indemnification could be imense to the affected utility. Impacts on public confidence are unknown. 1 Tentative Agenda and Schedule: The staff will prepare a report to the Congress providing an assessment of the impact of revised source terms on the Price-Anderson limit of liability. The report will be completed in late FY86 or early FY87, well before the expiration of the present act.
- The new source tenn information will be used in preparing Draft Environmental
- Statements beginning in late FY85.
1 + l
, i t
., ~. - _
.s., . .
?.; 17
.s .
~
- j '
t 5. AIR FILTRATION AND OTHER FISSION PRODUCT ATTENUATION SYSTEMS:
,j,; Background and Current Practice: Engineered safety features provided to -
ff mitigate accidental radioactive releases are reviewed and tested primarily
for effectiveness against iodine. Their perfonnance requirements are generally derived from the TID-14844 assumptions. Guidance is provided II,1 through several Regulatory Guides such as 1.3,1.4 and 1.52, and the '$ Standard Review Plan.' Such systems' include (1) containment spray systems, L. ,
C (2) recirculating air filters within containments (3) control room i habitabilitysystemairfilters,and(4)filteredbuildingexhausts. Since accidentally released iodine is presently assumed to be predominant in the form of molecular vapor, such mitigative features contain either a spray additive or charcoal impregnant intended to optimize the retention of molecular iodine. Some filter systems are also used to mitigate the radiological consequences of normal releases covered by 10 CFR 50, Appendix 1. Such releases also include radioactive iodine for which charcoal ffiters are known to be effective for removal. These systems are ineffective in the mitigation of noble gas releases. . PWR spray systems employ additives, such as hydrazine or sodium hydroxide,
; to enhance removal of elemental iodine. These systems are credited with
[ such removal during safety reviews. BWR suppression pools and BWR
. containment sprays, which do not employ additive systems, presently are
, not credited with fission product' removal or retention even though they are capable of absorbing elemental iodine to some degree.
, ,g ' .- - . .
,;, y- -
. ._,c .
p./ 18 ti
- . .s ,
O. As an inducement for the design of rapidly acting systems, accidentally W released ioolne is assumed to occur simultaneously with accident initiation f'i
,:<.. and to leak into the environment untreated for the duration of any time necesssry for the filters to achieve full capacity. Often, large n
a fractions of the estimated off-site thyroid doses are due to the iodine 9:, releases postulated during the first few minutes of the assumed accident. {.,; [. . y Possibility for and Character of Change: Present fission product mitigation systems, as discussed above, are optimized for the removal and retention of [. elemental and organic iodine. Source term research indicates that iodine e accidentally released in core melt accidents will not appear for a number of minutes after accident initiation and is likely to be present primarily in aerosol ' form along with aerosols of other volatile fission products, as well. Further, less volatile fission products and other aerosols may be released later in an accident. The effectiveness of present spray and filter systems against aerosols is believed to be high, but the need for attenuation of high concentrations such as could arise from core / concrete interactions has yet to be assessed. ESF filter systems contain particulate filters specifically designed to remove aerosols, while spray systems are known to be effective in removal of such contaminants. The use of additives in the spray system add little to the already high removal of aerosols by spray systems. Such additives are potentially important, however, in post-accident pH control of water that accumulates in plant sumps. At present there also exists a plant
) operational disincentive to have the spray system automatically actuated, lest - a corrosive additive harm equipment or personnel in the event of an inadvertent
( .. . 19 ?
?
spcay actuation. The staff concludes, on the basis of the above factors, that M sto'f guidance for auttmatic injection of spray additives should be eliminated, eg 2, To effect this change the staff has proposed as draft Branch Technical Position y, in Enclosure 5. Use of spray additives should be retained for PWR's, however,
;,y-N.' and be factored into the emergency operating procedures for control of the N re-evolution of iodine frcm sump water, as well as for post-accident chemistry 9: control. Credit should also be extended to BWR sprays and suppression pools a for fission product removal or retention in safety analyses. Such credit will be identified in a modification to the Standard Review Plan.
Since the effectiveness of present filter systems against all but high density mixes of aerosols is believed to be very high and the same filter systems
, provide for fission product attentuation in accidents of lessor consequences in' 4
which iodine may be present, no short-tenn changes with regard to existing filter systems are anticipated. Changes in filter system criteria that are
. proposed with future research are to be incorporated in Reg. Guide 1.52 along with other changes since the last revision of the Guide. High density aerosols and filter system bypass will be considered in the new revision. Backfitting for accidentally released elemental iodine removal is not to be required until
, such time as guidance for conditions covering the range of accidents from those involving coolant to fuel melting are fully assessed (see Enclosure 4). Cost and Benefits of Change: A large indirect benefit to most PWR licensees is
.- anticipated as a result of discontinuance of automatic injection of the spray additives. This is due primarily to reduced concern regarding the adverse
c_ . - :.. . .,w. .
..n. .
e,
.:?
20
.4 tg , > '. s Q:.',- effects of an inadvertent spray system actuation. A moderate, indirect benefit
>f ,3
.p. to BWR licenseet is anticipated as a result of the granting of fission produc. .:p is,j removal credit for BWR suppression pools and sprays. Moderate staff costs fi are expected to accumulate for changes to regulatory guide and the g,j .R , Standard Review Plan.
g. L.':t m,l } Tentative Agenda and Schedule: These changes are expected to be initiated in FY85 and completed in FY86. O h i. T w Y k n' . r 1 A 1 L M 4 i 8 9 e
- l, f
i 1
r . . .. .. v - t i. b,f 21 % 6. ACCIDENT MONITORING & MANAGEMENT ONSITE AND OFFSITE INSTRUMENTATION: y
*3 :;
j, Background and Current Practice: All plants are required to provide jf instrumentation to monitor plant variables and systems during and following
- anaccident(TMIActionPlanII.F.1.2). This instrumentation is qualified I' to be capable of operating under conditions estimated from the fission product 2,5 assumptions of TID-14844 and the peak accident temperature and pressure
.[ profiles. Under Regu'latory Guide 1.97, some instruments considered necessary to follow reactor accidents are qualified to a much harsher radiation - environment. Furthermore, research on accident management may improve guidance for prevention and mitigation. Instrumentation is also required for off-site monitoring of both routine . and accidental releases. Off-site instrumentation includes fixed thenno-luminescent detectors (TLD) at an array of locations surrounding the site, portable radiation measuring equipment, and portable air samplers. Possibility for and Character of Change: The instrumentation required to assess plant conditions during and following an accident is qualified to an environ-ment consisting of the non-mechanistic radiation doses indicated by TID-14844 release assumptions (except for a small nuder, as indicated above, which are
?.. qualified to a more stringent environment) combined with temperature and pressure profiles calculated for pipe break accidents not involving a core =I )3 melt.
4
- g. _..
. .m. . . _ . , ones, largely in the area of increased public confidence.
j Tentative Agenda and Schedule: The revised FRPCC policy statement on KI is
.. expected to be considered in the very near future. Assessment of proposed a ESF's is being pursued as part of the Severe Accident Research Program and standard plant reviews. Together with staff interaction with the industry
[ group (IDCOR),evaluationsexpectedinFY86andFY87areexpectedto demonstrate reduced risks and lowered requirements for additional ESF's for i standard plants. s
L ,. , 7
- 27 M
9 g. 4
- 8. SAFETY ISSUE EVa;3ATIONS ji
.3 D'
Background and Current Practice: Each newly raised generic safety issue is f'A In the first step, the staff assessed by the staff in a two-step process. establishes the ';;otential importance of the new safety issue relative to all others, in order to assign it a priority in the competition for limited
- 2, resources. This prioritization is a systematic application of PRA methods to 1
- estimate two indices of Srfety importance:
(a) Risk Importance - an assessment of the increase in societal risk posed by the gt:neric issue as indicated by the estimated dose to the surrounding population out to 50 miles from the plant; and
~
(b) Value/ Impact - a meesure of the cost effectiveness of resolving the . safety issue, i.e., the ratio of the risk reduction to the total cost (to I, . industry and the NRC) involved in developing and implementing the mode of resolution.
~
The indices are currently evaluated using source terms derived from
'^
WASH-1400. Changes in the indices to include proposed safety goals are under
- consideration.
Ai As generic safety issues achieve priority in the use of staff resources, they M enter the second step of the process. A more detailed evaluation is r e t l l
,, i' . . . l 28 N;
r 4
- s. n*
perfonned, comparing risks unoer existing conditions with what they are is estimated to be under improved ,2nditioris brought about by increased regulatory oc'.; , requirements addressing the generic safety issue. Among the factors d!*/ , e Fj considered in this second step are the potential reduction in risk to the y. ,p. public and the potential impact on the radiological exposure to plant h'.Y personnel associated with the proposed new requirements or backfit. .,h currently, the probabilistic estimate of public risk uses source terms derived s Gr .. from WASH-1400. On the other hand, occupational exposures are estimated in a 'A: j relatively direct manner based on measured onsite radiation levels and expected times of exposure.
. Evaluations of engineered safety features, environmental qualifications, and
~
many emergency procedures and technical specifications a're based in large
- measure on stylized accident calculations derived from TID-14844 and related assumptions. These calculations are used in safety evaluations for 0, construction permits, operating licenses and licenses amendments. The calculational methodology is described in TID-14844, various Regulatory Guides and the Standard Review Plan. The methodology includes not only assumptions i .
of fission product releases for a number of design bases accidents, but other as'sumptions relating to the transport and deposition of releases and their biological uptake that have not been updated since their codification before 1975. All such assumptions were deterministically designed to be conservative Y. in combination, but not individually to envelop all conditions. M
.}
- - -=
d
+
.~ -
29 J.s v.~ Oi l Possibility for and Character of Change: Depending upon the accident sequence
.t.*
pj and type of. plant, the new source term methodology can be used to estimate N
.. different accident consequences and risks than those derived from WASH-1400.
V.. y[j Such estimates used in prioritizing generic issues will utilize the new s: research. However, because the overall methodology allows for a rough but 1.f. Iki, adequate screening of the issues, no reprioritization is planned. In nany
,'- cases involving backfitting considerations, the assessed value/ impact could be . . s '. N '
changed producing differing results. The staff anticipates revising the l 'c. indices by which the safety importance of generic issues are evaluated to P. agree with the insights gained from the recent source term research. These may be different for different plant types or groups. In view of the source term research, and related information developed since' TID-14844 was publ,ished in 1962, the staff will reassess it's methodology for
; assessing the radiological consequences of design basis accidents. Four alternatives will be considered as follows:
l-
- 1. Evaluate whether the existing methodology is sufficiently conservative to y maintain the status quo, including the use of TID-14844 assumptions, for 1
; the suite of design bases accidents presently assessed.
d f. ' - 2. Determine whether modest changes to present assumptions, such as the fission product release assumptions in TID-14844, can be restated for Dm, J. F all plants, or groups of plants, to provide a suitable level of safety for
30 dF
.j
{ia, future regulatory analyses. Such restatements can be made in the form
,fi of Regulatory Guide and Standard Review Plan ch nges.
- 3. Restate all design bases accident assumptions in one of two ways; (lC:
i: I. f: I[-~, a) -recast ' design basis accident assumptions deterministically using the
,3 same kinds of dose assessments as presently employed; or s ,.A ,
- w. .
~
., b) do away with design basis accident dose calculations in favor of performance criteria for engineered safety features.
Rule, Regulatory Guide and Standard Review Plan changes may be necessary under this alternative.
- 4. Establish a risk based evaluation methodology using maturing probabilistic risk assessment techniques. Rule, Regulatory Guide and Standard Review Plan changes may also be necessary under this alternative.
Costs and Benefits of Change: Moderate staff effort is expected to revise the evaluation methodology of generic safety issues using new source term informction. Benefits are expected to be high since the risk categorization of an issue will be in keeping with the latest research data and emphasize l{., areas of risk significance. This change nay result in a general risk 2.' reduction for many generic issues.
~ ___ - ,
.:. c -
~
4I 31 5 a e a.. 2 p ,, A modest effort is also expected to identify and implement changes to the piesent staff design basis accident evaluation methodology. Buefits could be L;:. modest. q,,. h' Tentative Agenda and Schedule: Revision of the methodology for evaluating
~i -
generic safety issues is scheduled to begin in FY 85 and to be completed in
- - late FY 85 or early FY 86. Identification of alternatives to the present designbasisaccidentmethodologyisexpectedtocompfetedinFY86and i' ".'.I r
implemented in FY 87. r l s 4 s I 1 I l l l l L . ,
- l.
32 o .
- 9. SITING
'l:
Background and Current Practice: Siting criteria in 10 CFR Part 100 include several tests in which it must be demonstrated that (a) the site possesses f
, certaincharacteristics;(b)theplant-sitecombinationmeetscertain
- criteria; and (c) that the site is located sufficiently far from population h centers (remotesiting). Each site must have an exclusion area within which anapplicanthasauthbritytodetermineallactivitie. Beyond the exclusion
;' area lies the low population zone (LPZ) which may contain residents, the total l number and density of which are such that there is a reasonable probability that protective measures could be taken in their behalf in the event of an accident. Finally, Part 100 requires that the distance from the reactor to the nearest population center (of about 25,000 or more residents) must be at least one and one-third times the distance to the outer boundary of the LPZ. ,
The distances to the exclusion area boundary, the LPZ outer boundary and the population center are not numerically fixed, but depend upon the plant characteristics, including its maximum full power fission product inventory, and the complement and performance of certain engineered safety features. To test whether the plant-site combination meets the requirements of Part 100, ' ~ i a hypothetical core melt accident (TID-14844) is postulated involving the instantaneous release of 100% of the core inventory of noble gases and 50% of a the iodines into the containment. Half the iodines released are assumed to plate-out upon interior surfaces, while the remaining 25% is available for leakage. The containment is assumed to remain intact, but is presumed to leak
,.4m,__ c. _ _ __ . , _ . . _ _
t . 33
- at its design basis leak rate. The performance of fission product mitigating qli features (e.g., sprays, filters)areassessedinastylizedfashiontoestimate thyroid and whole body doses unlikely to be exceeded in hypothetical lf I,f f. individuals located at the exclusion area boundary and the LPZ outer boundary J for specified time periods. The plant-site combination is determined to be
.: - acceptable if the calculated doses do not exceed the guideline values given in ) Part 100. It is clear that current licensing practice involves a close t /. coupling of the site and the plant design. Also, the thyroid dose calculation, driven by assumptions of release of elemental iodine, is usually the limiting dose in determining the acceptability of the exclusion area boundary, the low population zone, and perfonnance requirements of certain fission product mitigating features.
No dose calculation is perfonned for individuals located at the population center. However, in the statement of considerations accompanying the issue of Part 100, it was noted that the population center distance requirement was added to provide for protection against excessive accident exposure doses to people in large centers, since accidents greater than the hypothetical accident postulated for siting purposes was considered conceivable although highly improbable. This statement was recognition of the possibility of 5 accidents involving containment failure and of their importance in siting
~
considerations. Since publication of the Reactor Safety Study (WASH-1400) in 4 1975, there has been a general recognition that although probably small, public accident risks are dominated by such accident considerations.
34 These siting criteria, promulgated in 1962, led to a significant improvement 4] in fission product mitigating engineered safety features ,in the late 1960's f.; ' and early 1970's. In 1975, the staff in response to objections that 10 CFR
! Part 100 did not preclude sites with very small exclusion area boundaries or . -)
in relatively densely populated regions, proposed (Regulatory Guide 4.7) that -/cy exclusion area distances should be about 0.4 miles or greater and that the [- average population density for the circular region surrounding a site should s. have no more than about 500 people per square mile within a distance of 30
~
E miles. The effect of this guide is to suggest a minimum stand-off distance from the reactor sites to large population centers (for example, a reactor i
~
should be located about 25 miles away from a city of 1,000,000 people). Sites where these values are exceeded are not forbidden, but should be shown to possess superior features in other respects to offset the disadvantage of high population. 'About 90 percent of the 75 U.S. power reactor sites meet the density criteria of the Guide; those that do not were reviewed and approved prior to 1975. The criteria of Regulatory Guide 4.7 were intended to provide j i a reasonable degree of separation from large population centers while maintaining a good availability of land area for potential future sites. i However, no explicit consideration of severe accidents was employed. In1979,astaffevaluationofsitingpolicyandpractice(NUREG-0625) recomended, among other things, an explicit consideration of severe accidents
- in siting, and a decoupling between plant design and siting requirements. The l Comission initiated rulemaking in this area in 1980, but suspended it about a
' .- o 35 9
year later, pending a re-evaluation of accident source tenns as well as an h evaluation of the proposed safety goal. o [ Possibility for and Character of Changes: Changes in siting could come about
. j primarily in two areas. These are in (1) evaluation of certain design basis C accidents used in licensing evaluations and technical specifications, and y, (2)considerationofaccidentsmoreseverethanthedesignbasis.
. i 9
- (1) The present design basis accident postulated for testing combinations of ESF and site characteristics, including the evaluation of the performance of containment and other engineered safety features, may contain significant errors when examined in light of our present understanding.
Among i.hese are an undue attention to iodine, particularly in the elemental form, and neglect of other fission products of importance, such as cesium. The possibilities for change with regard to design basis accidents are as follows: il ' l ;! I a) Develop revised design basis accident assumptions for siting and I. evaluation procedures for engineered safety feature performance that I l are in accord with the evolving understanding of fission product
- behavior under degraded core conditions.
I i b) Eliminate design basis accident radiological evaluations related to site / plant adequacy entirely by specifying a minimum set of required engineered safety features together with their perfonnance criteria, l . ,
36 y f ,- plusaminimumsetofsitecharacteristics(e.g.,distancesto .l exclusion area boundary, LPZ and population center). [ - 3
- .s
']. ?
(2) With regard to accidents beyond the design basis, siting criteria S- presently contain no explicit considerations of severe accidents, as noted above. Comission policy,10 CFR Part 100, and present staff (- f practice (via Regulatory Guide 4.7) does encourage siting away from c a 1 , t densely populated centers, but the present population density criteria have no clear link to severe accident risk. It should be noted, however, that more than 20 site-specific Environmental Impact Statements (EIS) have been performed by the staff which have included 'an explicit discussion of severe accident risks. These have shown the risks to be low for all the sites analyzed. In. addition, the Comission specia.1 proceeding for the Indian Point site, the highest population density reactor site, also explicitly considered severe accident risks and concluded these to be low. Based upon evaluations of low accident risks , for present sites using WASH-1400 source terms, together with the fact that there appears to be little incentive for siting in more populous
~
areas, the staff anticipates that a revision of siting criteria employing new source terms would represent no major changes from present staff practice with respect to site selection by applicants. Costs and Benefits of Change: Revision of siting criteria is expected to require a low to moderate effort in regard to staff resources. Most of this would be in evaluating the risks associated with the proposed criteria and m 4 - 4 -4 F .
-, _. . .. +
.. + . . . . . . . . *e
?.
, ., I ,
, l
* \
~
, 37 h;h
- ,y alternative approaches, including an evaluation with regard to the Safety Goal S.i criteria.
k;h, Benefits in terms of direct regulatory relief are expected to be low, since, I as mentioned earlier, the staff anticipates no major departures from present ( > staff practice and there is little incentive for more populous siting. Intangible benefits in terms of contributing to the Connission's policy on r- r preapproval of plant ' sites and enhancement of public confidence are expected to be high, however. The benefits also include the potential for a more stable and efficient licensing process. Tentative Agenda and Schedule: Since rulemaking in this area had been initiated in 1980 and suspended 'about a year later, some of the technical-work, especially in the area of land availability, is considered to be.still applicable. It is estimated that rulemaking could be reactivated in FY 1986 and completed in late FY 1986 or early FY 1987. e e 9
+
3
' S 9
., . .. + . .. . ,, .. , .. ..
*^
., + .
38 ENCLOSURE 3 35] PRELIMINARY BENEFIT-COST
SUMMARY
OF. AREAS
.,.y) f .'
TARGETED FOR SOURCE TERM RELATED CHANGES CHANGE IN if.' REGULATORY IMP'EMENTATION '4 REGULATORY TARGET COSTS BENEFITS. REQUIREMENTS *
'.. AREA
^^
\ 1 IDCOR-NRC Staff Search TO BE DETERMINED
- t. ' For Risk
, Outliers
' 4 1:- 2 Containment Performance
,.: Near Term Low High D FY86 Future Unknown Unknown U U 3 Equipment Moderate Moderate Qualification to High to High U FY88 4 Emergency Planning Moderate High D FY86 5 Accident Consequences Unknown ** Unknown ** U FY86
& Indemnification FY86 6 Air Filtration Low High D
& Other Fission Product Attenu-ation Methods High FY86 7 Accident High U Monitoring &
Management Onsite
& Offsite Instru-mentation ~. 8 Offsite Con- FY87 tamination & Low High D Recovery ' FY86 &
9 Safety Issue Moderate High U FY87 Evaluations 10 Siting Hoderate High 0 FY87
- Increase (I), Decrease (D), Unknown (U)
** Depends on Congress
*J ,,
T.,
-' ' E.E L A
' DRAFT
.. K T; BRANCH TECHNICAL POSITION AEB 6-4
':n CONTROL ROOM HABITABILITY pp i.i A.
BACKGROUND O:i Control room t.sbitabilty requirements historically have been dominated by considerations of thyroid doses that result from assumed accidental
,-{
j' exposure to gaseous iodine. Assumed releases of gaseous iodine have resulted in control room habitability system designs which incorporate such items as charcoal adsorbers in control room pressurization and
' [, recirculation systems. ,
o -. Recent source term research activities indicate that the accidental release of gaseous iodine as given in TID-14844* may be considerably overstated. That is, iodine releases from core melt accidents will primarily,be in the fom of cesium iodide as'an aerosol mixed with ae'rosols of other fission products. As a result the guidance for iodine protection as given in Standard Review Plan Section 6.4 may be unduly conservative with respect to iodine releases. Therefore, pending further decisions on the fission product mix and physical and chemical characteristics used in design basis accident assessments, the staff is suspending use of the thyroid criteria set forth in the Standard
' Review Plan for considerations which involve backfits of operating
" facilities. Suspension of such criteria will not apply to equipment
(: currently installed and maintained for which licensees request relief r,- from license commitments. In the interim the staff will utilize the whole body and beta skin dose guidelines of the Standard Review Plan in i detemining the acceptability of control room habitability systems.
- See Footnote 1 to 10 CFR 100. TID-14844 is Technical Information l .
Document-14844. "Calculation of Distance Factors for Power and Test Reactor Sites"; by DiNunno et al.; March, 1962. l
.,s I
* * ' + - - - - - - - - - - . . . - _ . , . _ _ _ _ , _ _ _ _ _ _
m w .
~
?, - ,,
+ '
_so
% -2
~c l2 Control rooms provide - .6ection to the operators by limiting the amount 5,f of inleakage to the cMtrol room envelope. The staff concludes the
'..'1 M. concept of a tight contrel room is still valid because the tighter the control room is, the more control there is over the amount of fission
~
ll,,;] products that reach the control room. Protection of the control rooms against aerosols, in turn, could be provided by designs that
}j minimize inleakage and/or incorporate HEPA filters on pressurization
- systems. Most current control room habitability systems incorporate HEPA filters and, therefore, have capability for removal of aerosols and operator protection.
S. BRANCH TECHNICAL POSITION For considerations which involve backfits of. operating facilities, control room habitability systems will be evaluated on the basis of estimated whole body and beta skin doses until additional perfortnance requirements are stated. These doses will be evaluated considering a TID 14844 release of 100% of the core noble gas inventory to the containment. . Control room designs should utilize low leakage design considerations f. E such as bubble tight damper construction, etc. Pressurization systems i should be installed with the capability of pressurizing the control room to 1/8" water gauge with a minimum amount of flow. Any pressurization or recirculation system backfitting should provide for (but not l-L i - necessarily include) the potential for future use of HEPA filters designed to remove aerosols. I i ,.i. .
7-a
- g'/ . E.MC L . 5
-e DRAFT BRANCH TECHNICAL ISITION AEB 6.5.2 INITIATION OF CONTAINMENT SPRAY CHEMIC'AL ADDITION A. BACKGROUND Design and operational requirements for chemical addition to containment sprays have been based on the TID-14844 assumptions regarding the amount and chemical form of accidentally released iodine fission products. TID 14844 postulates an instantaneous release of 50% of the core inventory of iodines to the containment. A substantial portion of this iodine was assumed to be gaseous iodine. The ability of spray systems using boric acid to remove elemental iodine has been assumed to be greatly enhanced by the addition of certain caustic chemicals.. The chemical addition was considered necessary very early in an assumed accident in order to
- reduce the offsite consequences of postulated releases of iodine from the containment through various leakage paths. Most PWRs provide such systems, BWRs do not. Without the hasty and periodic addition of the chemicals, projected offsite consequences for the design basis LOCA could exceed the dose guidelines specified in 10 CFR Part 100.
Recent source term research indicate that the initial accidental release of gaseous iodine as given in TID-14844 may be incorrect. That is, i iodine released from core melt accidents will primarily be in the fann of cesium iodide as an aerosol mixed with aerosols of other materials. Containment sprays without chemical addition could be demonstrated to be just as effective for removal of aerosols as sprays with the chemical
' * ** ", t a
a f' . ? .
. j ,, s '
n; 2 [ additives. The recent source term research t- not,' however, yet fully
'l r
addressed gaseous iodine releases from the sump later if such an . f.
; ,., , accident. Further, the release of such fissiin products is not g instantaneous (as postulated in TID-14844) but occurs at a later time in "O a core melt sequence. Consequently, imediate chemical addition to the
( ," ( ', contairiment sprays appears to be unnecessary.
' i, Many utilities have objected to designs that have no time delay in systems providing the addition of caustic chemicals to the containment ,
sprays because the possibility of inadvertent operation of the spray systems not only involves a personnel hazard, but potentially a very costly cleanup and plant damage. Indeed, the recent steam generator repairs at Three Mile Island, Unit 1, may be attributed to the automatic injection guideline. B. BRANCH TECHNICAL POSITION The imediate addition of caustic chemicals to the containment spray system are not required for plants so equipped. Provisions for manually initiated chemical system, however, should be provided for the addition 1 of such chemicals into containment spray systems in the event that the i the release of certain fission products dictates their use. The use of L
- the system should be based on a post accident sampling of the containment environment.
e i 4 L : .}}