ML20206L501
| ML20206L501 | |
| Person / Time | |
|---|---|
| Issue date: | 05/05/1999 |
| From: | Lavie S NRC (Affiliation Not Assigned) |
| To: | Barrett R NRC (Affiliation Not Assigned) |
| References | |
| PROJECT-689 NUDOCS 9905140042 | |
| Download: ML20206L501 (54) | |
Text
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION gg WASHINGTON, D.C. ensam mmg
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. May 5,1999 MEMORANDUM TO: Richard J. Barrett, Chief Probabilistic Safety Assessment Branch Division of Systems Safety and Analysis Office of Nuclear Reactor Regulation.
FROM:
- Stephen F.~ LaVie Probabilistic Safety Assessment Branch l
Division of Systems Safety and Analy Office of Nuclear Reactor Regulation
SUBJECT:
SUMMARY
OF APRIL 20,1999 MEETING WITH THE NUCLEAR ENERGY INSTITUTE (NEI) REGARDING STATUS OF STAFF EFFORTS TO ALLOW VOLUNTARY IMPLEMENTATION OF THE REVISED SOURCE TERM AT OPERATING REACTORS On April 20,1999, representatives of the Nuclear Energy Institute (NEI), various utilities, and vendors met with representatives of the Nuclear Regulatory Commission (NRC) at the NRC's offices in Rockville, Maryland. Attachment 1 provides a list of meeting attendees. The purpose of the meeting was to provide an opportunity for the NRC staff to update the members of the NEl task force on the implementation of the revised (NUREG 1465) source term at operating reactors. Attachment 2 is a copy of the meeting agenda. Because of a larger number of attendees than anticipated, the meeting was relocated from the location identified in the meeting notice to the Commissioner's hearing room.
Following introductions, the staff began with a brief presentation by Rich Barrett, Chief, Probabilistic Safety Assessment Branch (SPSB), on the recent changes to the NRR organization that have reassigned the lead on the source term rulemaking to SPSB.
1 1 Steve LaVie, SPSB, gave a brief status report on the proposed rule. The proposed rule was published in the Federal Register on March 11,1999 and the public comment period ends on
/
May 21,1999. The comments will be resolved and the final rule will be presented to the
/
Commission by August 28,1999.
Jay Lee, SPSB, gave a status' report on the review of the pilot projects. The review of the pilot f
project submitted for the Perry facility is complete and the SER has been issued. Some of the more significant aspects of this review were discussed. The pilot for the Indian Point facility is currently on hold awaiting information from the utility. The pilot for the Grand Gulf facility is currently in progress. Part of the Grand Gulf project involves a proposed General Electric Owner's Group topical report on the timing of the onset of the gap release phase. This topical report is currently under review. Slides used in this presentation are attached.
Steve LaVie, SPSB, gave a presentation on the content of the working draft of the draft guide that will support the proposed rule. Meeting participants were cautioned that this draft did not represent the final position of the staff and that everything discussed was subject to change.
9905140042990N5 PDR REVOP ERONUMRC fMfg MN EM mww peg u
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- Slides used in this presentation are attached. - During the presentation there were numerous
~l questions and comments raised by the meeting participants. There were three aspects of the draft guide that stimulated a large amount of discussion:
There was extensive discussion on the need to re-perform analysis of equipment environmental qualification (EO). The NRC staff stated that the currently envisioned position will be that the licensee must evaluate all impacts of the proposed change, including EQ. The NRC staff discussed the findings of the 1998 Re-baselining study as they applied to this issue. That study found that previous analyses based on the current accident source term (TID 14844) would bound similar analyses performed with the revised source term for components exposed to the containment atmosphere. The study showed, however, that the TID 14844-based analyses would be bounding initially, but as time progressed, the increased cesium in the revised source term would result in increased doses.
The crossover point occurred at about 30 days in one case, but more than four months in the second case. The NRC staff stated that these impacts would need to be evaluated.
The possibility that sensitivity analyses could show that extensive recalculations were not necessary was addressed. The NRC staff advanced the preliminary position that, if the licensee was allowed to use the revised source term, the staff's expectation was that the revised source term would become the design basis for all radiological calculations in the future, including EQ. If the licensee was able to show by sensitivity analysis that the calculation did not need to be revised to obtain approval for the revised source term, the NRC staff.would expect that any future revision of such a bounding calculation would incorporate the revised source term. The NRC staff stated that this would be necessary in order to meet the Commission's stipulation that a clear, logical, consistent design basis is maintained.
The issue of the numeric value of the control room dose criterion was discussed.
The NRC staff indicated that the entire issue of control room habitability was
' being addressed in a separate initiative and that the staff was meeting with NEl in a public meeting on this issue on April 28,1999. The NRC staff acknowledged the need to reconsider the numeric value, but indicated that this was unlikely to occur as a part of the current rulemaking, because of timing constraints on the current rulemaking.
The NRC staff discussed the Commission's position on the use of the alternative source term for relaxation of emergency preparedness requirements. This 1
elicited some discussion. The NRC staff emphasized that the alternative source
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i i term, in of itself, would not provide adequate justification to support relaxing emergency preparedness requirements, but that licensees who wished to pursue such applications would need to file requests for exemptions from the applicable requirements of 10.CFR 50.47 and Appendix E to 10 CFR Part 50.
j The lead participants decided to schedule a follow-on meeting in the mid-May time frame.
Project No. 689 1
- Attachments: As stated cc w/ attachments: See next page DISTRIBUTION: See attached page l
DOCUMENT NAME: A:\\MTG4_20,WPD To receive a copy of this document, indicate in the box: "C" = Copy w/o attachment, 'E' = Copy w/ attachment, "N" = No copy OFC SPSB E
SC:SPSB E
PM:REXB E
NAME SLaVie N MReinhard S.MagruderD DATE 5/ 4/99 5/3 /99 5/ % /99 i
OFFICIAL RECORD COPY l
j
Nuclear Energy Institute Project No. 689 cc:- Mr. Ralph Beedle
- Ms. Lynnette Hendricks, Director Senior Vice President Plant Support and Chief Nuclear Officer Nuclear Energy Institute Nuclear Energy Institute Suite 400 Suite 400 1776 I Street, NW 1776 i Street, NW.
Washington, DC 20006-3708 Washington, DC 20006-3708 Mr. Alex Marion, Director Mr. Charles B. Brinkman, Director Programs Washington Operations Nuclear Energy Institute ABB-Combustion Engineering, Inc.
Suite 400 12300 Twinbrook Parkway, Suite 330 1776 i Street, NW Rockville, Maryland 20852 Washington, DC 20006-3708 Mr. David Modeen, Director Engineering _
f Nuclear Energy Institute Suite 400 1776 l Street, NW Washington, DC 20006-3708 Mr. Anthony Pietrangelo, Director Licensing Nuclear Energy Institute Suite 400 1776 l Street, NW Washington, DC 20006-3708 Mr. Nicholas J. Uparuto, Manager Nuclear Safety and Regulatory Activities Nuclear and Advanced Technology Division Westinghouse Electric Corporation P.O. Box 355 -
Pittsburgh, Pennsylvania 15230 i
Mr. Jim Davis, Director Operations l
Nuclear Energy Institute Suite 400 l
17761 Street, NW Washington, DC 20006-3708
Distribution: Mtg Summary w/ NEl re Alternative Source Term Dated May 5, 1999 FILE CENTER j
Hard Copy PUBLIC SPSB R/F j
~OGC ACRS
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SMagruder SlaVie Email SCollins/RZimmerman BSheron WKane GHolahan TCollins RBarrett MReinhart SLaVie JLee SMagruder CTinkler, RES JSchaperow, RES MBlumberg
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REmch JHayes BWetzel MHart
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i NEl/NRC Meeting on Alternative Source Term April 20,1999
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List of Attendees j
i Name Oraanization J
Lenny Azzarello Duke Power Co.
Beth Wetzcl NRR/DPLM/PD3 i
Altheia Wyche Bechtel Gary Darden Virginia Power Bill Hopkins.
Bechetl Charles Jackson Con Edison David A.' Matchick Pa. Power & Light James L. Grover Westinghouse Electric Terry Heames ITSC Gopal Patel NUCORE John Duffy PSE & G Sreela Ferguson Stone & Webster Mike Withrow Entergy Greg Broadbent Entergy Michelle Hart NRR/DSSNSPSB John C. Nagle PSE & G PaulT. Reichert Raytheon Nuclear Rich Emch NRR/DLPM Chuck Nelson Alliant Energy Mike Cambria Applied Analysis Corp Jaun M. Cargas Applied Analysis Corp Tom Mscisz PECO Energy Jerry Burford Entergy Jack Hayes NRR/SPSB Bill Horin Winston & Strawn Mike Rabvaasky GPU Nuclear Jeff Robertson Numeric Applications Mark Blumberg '
NRR Rich Barrett NRR/DSSNSPSB Mark' Reinhart NRR/DSSNSPSB Jay Lee NRR/DSSNSPSB Steve LaVie NRR/DSSNSPSB Dave Leaver Polestar Kurt Cozens NEl Charlie Tinkler RES Jason Schaperow RES
- f.
Meeting Status of NRC Activities on implementation of Revised Source Term at Operating Reactors April 20,1999 Agenda introductions S. LaVie K. Cozens
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NRC Re-organization R. Barrett
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i
-i Rule Status S. LaVie
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Pilot Project Reviews J. Lee Highlights of Draft Guide S. LaVie Open Discussion all l
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Alternative Radiological Source Terms for Evaluating the Radiological Consequences of Design Basis Accidents at Boiling and Pressurized Water Reactors pl
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DG-1081 C
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Stephen F. LaVie NRR/DSSA/SPSB sSgnrc. gov 301 415 1081 P.ELIMINA.,!!iPO.MA,80N-SUBJEC, TO CNAN.E
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,,,, 4 Draft Guide Outline A. INTRODUCTION B. DISCUSSION C. REGULATORY POSITION
- 1. Implementation of Altemative Source Term 1.1 Generic Considerations Mf_M i
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1,2 Scope ofimplementation 1.3 Scope of Required Analyses 1.4 Riskimplications
/h 1.5 Submittal Requirements
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1.6 FSAR Requirements
.........,.......,,..-.s..,,.........
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a Draft Guide Outline C. REGULATORY POSITION (Con 1)
- 2. Attributes of an Acceptable Alternative Source Term
- 3. Accident Source Term 3.1 Core inventory 3.2 Release Frac #ons 3.3 Timing of Release Phases 3.4 Radionuclide Compool#on 3.5 Chemical Form 3.6 Fuel Damage in Non4.OCA DBAs P
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.....c.
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Draft Guide Outline C. REGULATORY POSITION (Con 1)
- 4. Dose Calculational Methodology 4.1 Offsite Dose Consequences 4.2 Control Room Dose Consequences 4.3 AcceptanceCriteria
- 5. Analysis Assumptions and Methodology
'5.1 GeneralConsiderations Aa iyok oueury creer Ar sneinewat seksuanremennes A isrvnene eruu,nerk arpur wuse ApPnessuny erPrior LkenehnBeak 5.2 Accident Specific Assumptions
[W 5.3 Meteorology Assumptions
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Draft Guide Outline D. IMPLEMENTATION E. REGULATORYIMPACT ANALYSIS Appendices A.
Assump# ens - DBA LoCA B. AssumpWons - FHA C.
AssumpWons - PWR sGTR l
G.
AssumpWons - PWR REA
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J L AssumpWons - EQ
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Generic Considerations
- The NRC staff will allow licensees to pursue technicallyjustifiable uses of the ASTs in the most flexible manner compatible with maintaining a clear, logical, and consistent design basis.
I
- The NRC staff expects to approve these license amendment requests if the facility, as modified, will i
continue to provide sufficient margin of safety with adequate defense-in-depth to address unanticipated l
events and to compensate for uncertainties in accident progression and analysis assumptions and parameter inputs.
21.
3
1 Margin of Safety
- margin of safety = the difference between calculated parameters (e.g., dose) and the associated regulatorylimit.
- The margin of safety is a function of specific values in the T/S and other values, such as assumed accident initial conditions.
Iso.se rus.maung may have erreco N
- _),
Defense-in-Depth
- Defense in depth = system redundancy, indepen.
dance, and diversity preserved commensurate with the frequency & consequences of challenges to the system
- Compliance with the GDCs
- Removal of previously required ESF equipment should not invalidate assumptions made in facility PRAs or adversely impact the facility's severe accident management program
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- (M)_
4
Design Basis Integrity
- Selective implementations allowed
- May be able to use sensitivity analyses to show old analysis is still bounding
- Could have two tiers of analyses-old source term /
new source term soulon bases mustbe clear j
eFuture revisions of strese analyses should utilise the updated usumpeans.
- Superceded analyses should be removed (W) 1
- \\ ;_.,
- EPP planning basis involves of a spectrum of DBA and severe accidents.
- AST has been tailored specifically for DBA analysis use.
- AST is not representative of the wide spectrum of possible events that comprise the planning basis and, therefore, is insufficient by itself as a basis for requesting relief from EPP requirements.
- (O M)
...,..,..e....,,.......,.....
a 5
1 Scope of implementation
- Selective implementation:
i
- One or more (but less than all) characteristics of AST addressed or/and
- Limited subset of DB radiological analyses upgraded
- Maximum flexibility
- Technicaljustifiable
- Clear. consistent, design basis nN):,,
Scope of Implementation (Con't)
- lf full LOCA dose analysis is done:
- 50.59 basis of future changes (unless USQ)
- lf full LOCA dose analysis is not done:
- SER will be limited to the specific selective application
- Must re apply under 50.67 to extend AST/TEDE to a different application.
6
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l Scope of Required Analyses
- Should evaluate all of the radiological and non-radiological impacts of the proposed actions l
- Scope is a function ".# specific modification
- Complete re-calcuietion not necessary-update those analyses affected Alf DB analyses are re calculated, all affected assumptions and inputs should be updated, all characteristics of AST & TEDE em
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Scope of Req'd Analyses (Con't)
- Two categories of impacts
- Due to plant modification
- Due to AST characteristics
- Staff did re baselining study (SECY-98-154)
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Scope of Req'd Analyses (Con't)
- Cross over point is 30 days to 4 months (maybe loneer)
- Need to address this impact in analyses with >30 day exposure
- Need to address all radiological requirements in regs
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and commitments in the submittal text--Does not imply calculete ell.
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Risk Implications
- No increase in risk due to AST ltself
- Plant modification could have risk impact (1) Elect to use PRA info to support modification
- Regulatory Gutde 1.174 (2)If PRA exists
- Need to evaluate if modification invalidates PRA assumptions (3} if no PRA, but retiring ESF equipment
- Need to consider risk impact
- NRC may request risk info if adequate protection is questioned
- Control room likely to be special case B
Accident Source Term
- Applicable to LWR fuel with peak burnup
<62,000 MWD /MTU
- Potential change for RIAs
- Foreign Tests
{
- RES efforts in progress j
- MOX not addressed r%
(W)
)
i Core Inventory
- Current licensed fuel parameters:
. Fuel enrichment
- Fuel burn-up
- 1.02x rated power
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9
Release Fractions
- For LOCA
- Gap release & early in vessel phases only
- From NUREG-1465
- Non-LOCA Gap Fractions
- 12% I 131
- 15% Kr-85
- 10% other NG,1
- 10% alkall metals
/%
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Timing / Radionuclides ORCS activity phase deleted
- except purge case
- Gap & early in-vessel same as NUREG-1465
- Provision to re-calculate onset of gap phase
- Radionuclides same as NUREG-1465 F%
(W).
10 I
Chemical Form
- Release to CNMT
+95% Csl I
- 4.86% elemental j
- 0.15% organic f
- 100% elemental iM) y,,,,, p PRSLIMINAR,IN..RMA,0.N-8WSJ.C,19 CNAN m
Fuel Damage in Non-LOCA
- Need to Consider
- Fuel clad breach
- Fuel that reaches melt temp
~ Can use approved enthalpy methods in lieu of DNBR
- But, not for RDA, REA (OW)
...M..,......,,..8...,,.....
11
Dose Calculation Methodology
- Control Room
- Same accident progression as offsite analysis
- Lkeless norHumservedve hw CRM
- Credit for ESFs
- Generally, no credit for PPE or Kl
- Offsite
- TEDE, worst 241our at EAB
- M*)
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Acceptance Criteria
- Same "small fraction" SRP criteria rhh
- .)..
12
7 l
General Analysis Assumptions l
l
- Subject to Appendix B, Part 50
- Credit for ESFs if
- Safety-grade
- Emergency power
- No single failure
- Auto actuation (some exceptions)
- General guidance on parameter values iN) l General Analysis Assumptions (con't)
- Prior licensing basis
- New or unreviewed issues may be created by application that warrants review against current staff positions f%
l
- h)
Accident Specific Assumptions
- Provided in appendices
- Existence as appendix does notimplyrequirement
- DBA are not real accidents osurrogates intended to test one or more aspects of plant design e Conservative in order to address uncertainties in accident progr.asion
- Attempts made to change assumptions, where data available estaffis interested in su90cstions for other supportable changes (nW)
...a. ~..,,..
4 Meteorology Assumptions
- (1) Use current approved values in FSAR
- Ensure applicability to release paths / accidents
- (2) Recalculate
- Offsite: RG 1.145 (e.g. PAVAN code)
- CR: Murphy-Campe or ARCON-98*
eData collection per FSAR / TS / RG1.23
- aome covests (M) 14
WORKING DRAFT -SUBJECT TO CHANGE Appendix A'
)
1 ASSUMPTIONS USED FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A LWR LOSS OF COOLANT ACCIDhNT This appendix provides assumptions acceptable to the staff for evaluating the radiological consequences of a loss of coolant accident at light water reactors. These assumptions supplement the
-_ guidance provided in the main body of this guide.
Appendix A, Genaral Design Criteria for Nuclear Power Plants, to 10 CFR Part 50 defines LOCAs as those accidents that result in loss of coolant inventory at rates that exceeds the capability of the reactor coolant makeup system. Included are leaks up to a double-ended rupture of the largest pipe of the reactor coolant system. The LOCA, as with all DBAs, is a conservative surrogate accident that is intended to challenge selective aspects of the facility design. Analyses are perforined using a spectrum of break sizes to evaluate fuel and ECCS performance. With regard to radiological consequences, a large break LOCA is assumed as the design basis case so as to evaluate the performance of release mitigation systems and the containment and to evaluate the proposed siting of the facility. The assumption of a large break LOCA maximizes the source term available for release. However, some safety systems, such as containment isolation, containment sprays, and control room isolation, may not actuate or may be delayed for LOCAs involving smaller breaks. Although the large break LOCA is the radiological design basis case, individual case consideration may need to be given to evaluating small break LOCAs where delays in assumed system response may result in increased radiation doses. In these small break LOCA analyses the data of Tables 1 and 2 should be adjusted based on suitably conservative projections of the fuel damage. If no fuel melt is projected, the gap fractions of Table 3 should be assumed.
Source Term 1.
Acceptable assumptions regarding core inventory and the release of radionuclides from the fuel are provided in Section C.3 of this guide.
2.
If the sump or suppression pool pH is controlled at values of 7 or greater, the chemical form of radiciodine released to the containment should be assumed to be 95% cesium iodide (Csl),4.85 percent elemental iodine, and 0.15 percent organic iodide lodine species for sump or suppression pool pH values less than 7 will be evaluated on an individual case basis. Evaluations of pH should consider the effect of acids created during the LOCA event, e.g., pyrolysis and radiolysis products. With the exception of elemental and organic iodine and noble gases, fission products should be assumed to be in particulate form.
Transport in Primary Containment l
3.
The assumptions related to the transport, reduction, and release of radioactive material in and from i
the primary containment in PWRs or the drywellin BWRs are as follows:
a.
The radioactivity released from the fuel should be assumed to mix instantaneously and I
I homogeneously throughout the free air volume of the primary containment in PWRs or the drywellin BWRs as it is released. This distribution should be adjusted if there are internal compartments that ha ie limited ventilation exchange. The suppression pool free air volume may be included provided there is a mechanism to assure mixing between the drywell to the wetwell. The release into the containment or drywell should bc assumed to terminate at the end of the early in-vessel phase.
~
WORKING DRAFT -- SUBJECT TO CHANGE l
WORKING DRAFT - SUBJECT TO CHANGE b.
Reduction in containment airbome radioactivity by natural deposition within the containment may be credited. Acceptable models for removal of iodine and aerosols are described in the Chapter 6.5.2, Containment Spray as a Fission Product Cleanup System, of the SRP i
and in NUREGICR-6189, A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments. The latter model is incorporated into the analysis code RADTRAD (Ref. <radtrad>). The prior practice of deterministically assuming a 50% plateout of iodine released from the fuel should not be done as this treatment of plateout is inconsistent with the characteristics of the revised source terms c.
Reduction in containment airbome radioactivity by containment spray systems that have been designed and are maintained in accordance with the guidance of Chapter 6.5.2 of the SRP may be credited. Acceptable models for removal of iodine and aerosols are described
~ in Chapter 6.5.2 of the SRP' and NUREGICR 5966, A Simplified Modelof Aerosol removal 2
by Containment Sprays. This simplified model is incorporated into the analysis code RADTRAD.
l The evaluation of the containment sprays should address areas within the primary containment that are not covered by the spray drops. The mixing rate attributed to natural convection between sprayed and unsprayed regions of the containment building, provided that adequato flow exists between these regions, is assumed to be two tumovers of the unsprayed region (s) per hour, unless other rates are justified. The containment building atmosphere may be considered a single, well-mixed volume if the spray covers at least 90%
of the volum'e and if adequate mixing of unsprayed compartments can be shown.
d.
Reduction in containment airbome radioactivity by in-containment recirculation filter systems may be credited if these systems meet the guidance of Regulatory Guide 1.52 and GL
< charcoal >. The filter media loading due to the increased aerosol release associated with the revised source term should be addressed.
e.
Reduction in containment airbome radioactivity by suppression pool scrubbing in BWRs should generally not be credited. However, the staff may consider such reduction on an individual case basis. The evaluation should consider the relative timing of the blowdown and the fission product release from the fuel, the force driving the release through the pool, and the potential for any bypass of the suppression pool. An acceptable model for removal of lodine and aerosols is described in NUREGICR-6153, A Simplified Model of Decontamination by BWR Steam Suppression Pools. Analyses should consider iodine re-evolution if the suppression poolliquid pH is not maintained greater than 7.
' The SRP establishes a maximum decontamination factor (DF) for elemental iodine based on the maximum lodine activity in the primary containment atmosphere when the sprays actuate divided by the activity of iodine remaining at some time after decontamination. Since the activity is now assumed to be released continuously over a period of time, the rnaximum activity needs to be redefined. If the release from the fuelis to be modeled as a linear ramp over the duration of the release phase, the maximum activity should be the activity remaining at the end of the early in-vessel phase, if the release from the fuel is assumed to occur as a step increase at the start of the early in-vessel release phase, maximum activity should be the activity assumed to be released at that time.
8 This document describes statistical formulations with differing levels of uncertainty. The removal rate constants selected for use in design basis calculations should be at the 90% percentile, i.e., the probability that the true value would be less is 10% or less. For BWRs, the simplified model should be used only if the release from the core is not directed through the suppression pool, lodine removal in the suppression pool affects the lodine species assumed to be initially present by the model.
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-WORKING DRAFT - SUBJECTTO CHANGE f.
Reduction in containment airbome radioactivity by retention in ice condensers, or other engineering safety features not addressed above, should be evaluated on an individual case basis. See SRP Chapter 6.5.4.
~
The primary containment should be assumed to leak at the peak pressure technical g.
specification leak rate for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and at 50% of this leak rate for the remaining duration of the accident. Leakage from subatmospheric containments is assumed to terminate when the containment is brought to and maintained at a subatmospheric condition as defined by technical specifications.
h.
For BWRs with Mark lil containments, the flow rate from the drywell into the primary containment should be based on the steaming rate
- of the heated reactor core, with no credit for core debris relocation. This flow rate should be assumed during the two-hour period between the initial blowdown and termination of the fuel radioactivity release (gap and early in-vessel release phases). After two hours, the radioactivity is assumed to be uniformly distributed throughout the drywell and the primary containment.
i.
If the primary containment is routinely purged during power operations, releases via the purge system prior to containment isolation should be analyzed and the resulting doses l
summed with the postuhted doses from other release paths. The purge release evaluation l
should assume that 100% of the radionuclide inventory in the RCS liquid is released to the j
containment at the initiation of the LOCA. This inventory should be based on the technical specification RCS equilibrium activity. lodine spikes need not be considered. If the purge system is not isolated prior to the onset of the gap release phase, the release fractions associated with the gap release and early in vessel phases should be considered as applicable.
Dual Containments 4.
For facilities with dual containment systems, the assumptions related to the transport, reduction, and release of radioactive materialin and from the secondary containment and/or enclosure buildings are as follows:
a.
Leakage from the primary containment should be considered to be collected, processed by ESF filters, if any, and released to the environment via the secondary containment exhaust system during periods in which the secondary containment has a negative pressure as defined in technical specifications. Credit for an elevated release should be assumed only if the release point is more than two and one-half times the height of any adjacent structure.
b.
Leakage from the primary containment is assumed to be released directly to the environment as a ground level release during any period in which the secondary containment does not have a negative pressure as defined in technical specifications.
c.
The effect of high wind speeds on the ability of the secondary containment to maintain a negative pressure should be evaluated on an individual case basis. Wind speed to be 8 Analyses performed with the MELCOR computer program for a BWR Type 6 facility with a Mark lil containment design estimate flow at 2800 cfm for large break LOCAs,28.000 cfm for small-break LOCA with ADS and 800 cfm for small-break LOCA without ADS.
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, assumed is that one hour average value that is exceeded only 5% of the total number of hours in the data set. Ambient temperatures used in these assessments should be the one
{
hour average value that is exceeded only 5% or 95%, whichever is conservative for the intended use, of the total numbers of hours in the data set.
l d.
Credit for dilution in the secondary containment may be allowed when adequate means to cause mixing can be demonstrated. Otherwise, the leakage from the primary containment j
should be assumed to be transported directly to exhaust systems without mixing. Credit for
)
mixing, if found to be appropriate, should generally be limited to 50%. This evaluation should consider; magnitude of containment leakage in relation to contiguous building volume and/or exhaust rate, location of exhaust plenums relative to projected release
)
locations, recirculation ventilation systems, intemal walls and floors that impede stream flow I
between the release and the exhaust.
e.
Prim'ary containment leakage that bypasses the secondary containment should be f
evaluated at the bypass leak rate incorporated in the technical specifications. If the bypass leakage is through water, e.g., via a filled piping run that is maintained full, credit for retention of iodine and aerosols may be assumed. This reduction should be evaluated on a case-by-case basis.
f.
Reduction in the amount of radioactive material release from the secondary containment due to ESF filter systems may be taken into account provided that these systems meet the guidance of Regulatory Guide 1.52 a..d GL < charcoal >.
ESF System Leakage i
5.
Engineered safety feature (ESF) systems that recirculate sump water outside of the primary containment are assumed to leak during their intended operation. This release source includes leakage through valve packing glands, pump shaft seals, flanged connections, and other similar components. This i
release source may also include leakage through valves isolating interfacing systems (Ref. <GL 9156>).
The radiological consequences from the postulated leakage should be analyzed and combined with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA. The following assumptions should be used in evaluating the consequences of leakage from engineered safety feature (ESF) components outside the primary containment for BWRs and PWRs:
a.
All of the fission products, with the exception of noble gases, released from the fuel should be assumed to be instantaneously and homogeneously mixed in the primary containment sump water (PWRs) or suppression pool (BWRs) at the time of release from the core. In lieu of this deterministic approach, suitably conservative mechanistic models that transport containment airbome activity to the sump water may be used. Note that many of the parameters that make spray and deposition models conservative with regard to containment airbome leakage are non-conservative with regard to the buildup of sump activity, b.
The leakage should be taken as two times the sum of the simultaneous leakage from all components in the ESF recirculation systems above which the technical specifications, or licensee commitments to item lli.D.1.1 of NUREG-0737, would require declaring such systems inoperable. The leakage should be assumed to occur, starting at the earliest time that the recirculation flow occurs in these systems and ending at the latest time that the releases from these systems are terminated. Consideration should also be given to design WORKING DRAFT -- SUBJECT TO CHANGE
i WORKING DRAFT --- SUBJECT TO CHANGE leakage through valves isolating ESF recirculation systems from tanks vented to
. atmosphere, e.g., ECCS pump miniflow retum to refueling water storage tank.
i c.
In addition to the leakage specified in paragraph 5.b, the evaluation should assume leakage
^
from a gross failure of a passive component at the rate of 50 gallons per minute, starting at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the accident and lasting for 30 minutes. This evaluation is not required if the facility design provides an ESF ventilation filtration system that exhausts the areas of potential leakage.-
d.
With the exception of iodine, all radioactive materials in the recirculating liquid should be assumed to be retained in the liquid phase.
e.
If the temperature of the leakage exceeds 212 'F, the fraction of iodine in the liquid that becomes airbome should be assumed equal to the fraction of the leakage that flashes to vapor. This flash fraction, FF, should be determined using a constant enthalpy, h, process, based on the maximum time-dependent temperature of the sump water circulating outside the containment:
pp, k N N
Where: hn is the enthalpy of liquid at system design temperature and pressure; ha is the enthalpy of liquid at saturation conditions (14.7 psia,212'F); and h,, is the heat of vaporization at 212*F If the temperature of the leakage is less than 212 'F or the calculated flash fraction is less than 10%, the amount of iodine that becomes airbome should be assumed to be 10% of the activity in the leaked fluid, unless a smaller amount is justified based on the actual sump pH history and area ventilation rates.
a.
The radioiodine that is postulated to become airborne should be assumed 100% elemental and should be assumed to be released to the environment. Reduction in release activity by dilution or holdup within buildings, or by ESF ventilation filtration systems may be credited where applicable. Filter systems used in these applications should be evaluated against the guidance of Regulatory Guide 1.52 and GL <chucos.
MSlV Leakage in BWRs -
j 6.
For BWRE, (he main steam isolation valves (MSIVs) have design leakage that may result in a radioactivity release. The radiological consequences from postulated MSIV leakage should be analyzed and combined with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA. The following assumptions should be used for evaluating the consequences of MSIV 'anksga:
a.
For the purposes of this analysis, tiie activity available for release via MSIV leakage should be assumed to be that activity determined to be in the drywell for evaluating containment leakage. No credit should be assumed for activity reduction by the steam separators or by lodine partitioning in the reactor vessel.
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WORKING DRAFT - SUBJECT TO CHANGE b.
All of the MSIVs should be assumed to leak at the maximum leak rate above which the technical specifications would require declaring the MSIVs inoperable. The leakage should be assumed to continue for the duration of the accident.
c.
Reduction of the amount of released radioactivity by deposition and plateout on steam system piping upstream of the outboard M31Vs may be credited, but the amount of reduction in concentration allowed will be evaluated on an individual case basis. Acceptable models for removal of lodine and aerosols are described in References <NR6604>, <NR6153>
and <kline>. The model should generally be based on the assumption of well-mixed volumes rather than on slug flow, but other models may be used if justified.
d.
In the absence of release collection and treatment by ESFs such as MSIVLCS, or as described in paragraph 6.e below, the MSIV leakage should be assumed to be released to the environment as an unprocessed, ground level release. Holdup and dilution in the turbine building should not be assumed.
e.
Reduction in MSIV releases due to holdup and deposition in main steam piping downstream of the MSIVs and in the main condenser, including treatment of air ejector effluent by offgas i
systems, may be credited if the components and piping systems used in the release path are capable of performing their safety function during and folicwing a safe shutdown 1
earthquake (SSE). The amount of reduction allowed will be avaluated on an individual case basis. See paragraph 6.c for guidance on acceptable models.
Containment Purging 7.
The radiological consequences from post-LOCA primary containment purging as a combustible gas
' or pressure control measure should be analyzed. If the installed containment purging capabilities are maintained for purposes of severe accident management and are not credited in any design basis analysis, radiological consequences need not be evaluated. If applicable, the results of this analysis should be combined with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA.
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WORKING DRAFT - SUBJECT TO CHANGE Appendix B ASSUMPTIONS USED FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT l
This appendix provides assumptions acceptable to the staff for evaluating the radiological j
consequences of a fuel handling accident at light water reactors. These assumptions supplement the guidance provided in the main body of this guide.
i Source Term 1.
Acceptable assumptions regarding core inventory and the release of radionuclides from the fuel are provided in Section C.3 of this guide. In addition:
a.
The number of fuel rods damaged during the accident should be t,ased on a conservative analysis that considers the most limiting case. This analysis should consider parameters such as the weight of the dropped heavy load, or the weight of a dropped fuel assembly (plus any attached handling grapples), the height of the drop, and the compression, torsion, and shear stresses on the irradiated fuel rods. Damage to adjacent fuel assemblies, when present (e.g., events over the reactor vessel), should be considered. Chapter 15.7.4, Radiological Consequences of FuelHandling Accidents, of the SRP provius an example of a conservative bounding analysis, b.
The gap activity fractions of Table 3 should be assumed. All of the gap activity in the damaged rods is assumed to be instantaneously released. Radionuclides that should be considered include xenons, kryptons, halogens, cesiums and rubidiums.
c.
Of the radiolodine released from the fuel,99.75 percent of the iodine released should be assumed to be in the form of elemental iodine and 0.25% organic species.
Release Transport 2.
If the depth of water above the damaged fuel is 23 feet or greater, the decontamination factors for the elemental and organic species are 500 and 1, respectively, giving an overall effective decontamination factor of 200 (i.e.,99.5% of the totaliodine released from the damaged rods is retained by the water). This difference in decontamination factors for elemental and organic iodine species results in the iodine above the water being composed of 44% elemental and 55% organic species. If the depth of water is not 23 feet, the decontamination factor will have to be determined on a case-by-case method (Ref. < burley >).
3.
The retention of noble gases in the water in the fuel pool or reactor cavity is negligible (i.e.,
decontamination factor of 1). Particulate radionuclides are assumed to be retained by the water in the fuel pool or reactor cavity (i.e., infinite decontamination factor).
4.
For fuel handling accidents postulated to occur within the fuel building, the following assumptions should be made:
a.
The radioactive material that escapes from the fuel pool to the fuel building is assumed to be released to the environment over a two hour time period.
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WORKING DRAFT - SUBJECT TO CHANGE b.
Reduction in the amount of radioactive material release from the fuel pool due to ESF filter systems may be taken into account provided that these systems meet the guidance of Regulatory Guide 1.52 and GL < charcoal >. Delays in radiation detection, actuation of the ESF filtration system and/or diversion of ventilation flow to the ESF filtration system'should
~~~
' be determined and accounted for in the radioactivity release analyses.
c.
The radioactivity release from the fuel pool should be assumed to be drawn into the ESF filtration system without mixing or dilution in the fuel building.
5.
For fuel handling accidents postulated to occur within the containment, the following assumptions should be made:
a.
If the containment is isolated during fuel handling operations, no radiological consequences need to be analyzed.
b.
If the containment is open during fuel handling operations, but designed to automatically
)
isolate in the event of FHA, the release duration should be based on delays in radiation detection and completion of containment isolation. If it can be shown that containment isolation occurs prior to the release of radioactivity to the environment', no radiological consequences need to be analyzed.
c.
If the containment is open during fuel handling operations (e.g., personnel air lock or equipment hatch open)2 the radioactive material that escapes from the reactor cavity pool to the containment is released to the environment over a two hour time period.
d.
. Reduction in the amount of radioactive material released from the containment by ESF filter systems may be taken into account provided that these systems meet the guidance of Regulatory Guide 1.52 and GL < charcoal >. Delays in radiation deteClion, actuation of the ESF filtration system and/or diversion of ventilation flow to the ESF filtration system should be determined' and accounted for in the radioactivity release analyses, e.
Credit for dilution or mixing, of the activity released from the reactor cavity, due to natural or forced convection inside the containment may be considered on a case-by-case basis.
Such credit is generally limited to 50% of the containment free volume.
' This analysis should consider the time for the radioactivity concentration to reach levels corresponding to the monitor setpoint, instrument line sampling time, detector response time, diversion damper alignment time and filter system actuation, as applicable.
2 The NRC staff will require administrative controls be established to close the airiock or hatch in less than 30 minutes. These controls will generally require a dedicated individual be present at the open airlock or hatch while fuel handling operations are in progress and that this individual have any necessary equipment to close the airlock or hatch in the required time. Radiological analyses should generally not credit this manual isolation.
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~
Appendix C ASSUMPTIONS USED FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A BWR ROD DROP ACCIDENT This appendix provides assumptions acceptable to the staff for evaluating the radiological consequences of a rod drop accident at BWR light water reactors. These assumptions supplement the guidance provided in the main body of this guide.
.1.
Acceptable assumptions regarding core inventory and the release of radionuclides from the fuel are provided in Section C.3 of this regulatory guide. The release from breeched fuel clad should be based on the gap inventory fractions in Table 3 and the estimate of clad damage. The release from melted fuel should be based on the early in-vessel phase data in Table 1 and the percentage of the fuel affected.
2.
If no or minimal' fuel damage is postulated for the limiting event, the activity released should be the maximum coolant activity allowed by technical specification. The iodine concentration in the primary coolant is assumed to correspond to the following two cases in the NSSS vendor's standard technical specifications:
a.
the concentration that is the maximum value (typically 4 pCi/gm DE l-131) permitted and corresponds to the conditions of an assumed pre-accident spike; and b.
the concentration that is the maximum equilibrium value (typically 0.2 pCi/gm DE l-131) permitted for continued full power operation.
3.
The assumptions related to the transport, reduction, and release of radioactive material from the fuel and the reactor coolant are as follows:
a.
The activity released from the fuel from either the gap or from fuel melt is assumed to be instantaneously mixed in the reactor coolant within the pressure vessel.
b.
Credit should not be assumed for partitioning in the pressure vessel or for removal by the steam separators.
c.
Of the activity re! eased from the reactor coolant within pressure vessel,100% of the noble gases,10% of the iodine, and 1% of the remaining radionuclides are assumed to reach the turbine and condensers, d.
Of the activity that reaches the turbine and condenser,100% of the noble gases,10% of the iodine, and 1% of the particulate radionuclides are released to the environment. The turbine 2
and condensers leak to the atmosphere as a ground level release at a rate of 1% per. day
'..The activity assumed in the analysis should be based on the activity associated with the projected fuel damage or the maximum technical specification values, whichever maximizes the radiological consequences. In determining dose equivalent 1131 (DE l-131), only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel damage should nct be included.
2 If there are forced flow paths from the turbine or condenser, such as unisolated motor vacuum pumps, unprocessed air ejectors, etc., the leakage rate should be assumed to be the flow rate associated with the most limiting of these paths. Credit for collection and processing of releases, such as by off gas or SBGT, will be WORKING DRAFT - SUBJECT TO CHANGE
WORKING DRAFT -SUBJECT TO CHANGE for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, at which time the leakage is assumed to terminate. No credit should be assumed for dilution or holdup within the turbine building. Radioactive decay during holdup in the turbine and condenser may be assumed.
e.
In lieu of the transport assumptions provided in paragraphs b through d above, a more mechanistic analysis may be used on a case-by-case basis. Such analyses accounts for i
the quantity of contaminated steam carried from the pressure vessel to the turbine and condensers based on a review of the minimum transport time from the pressure vessel to the first main steam isolation (MSIV) and considers MSIV closure time.
j f.
The release fiom the reactor coolant within pressure vessel should be assumed to consist of 95% Csl as an aerosol,4.85% elemental, and 0.15% organic. The release from the l
turbine and condenser should be assumed to be 100% elemental.
1 l
l i
I considered on a case-by-case basis.
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WORKING DRAFT - SUBJECT TO CHANGE Appendix D ASSUMPTIONS USED FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A BWR MAIN STEAM LINE BREAK ACCIDENT
~
~This appendix provides assumptions acceptable to the staff for evaluating the radiological consequences of a main steam line accident at BWR light water reactors. These assumptions supplement the guidance provided in the main body of this guide.
Source Term 1.
Acceptable assumptions regarding core inventory and the release of radionuclides from the fuel are provided in Section C.3 of this regulatory guide. The release from breeched fuel clad should be based on the gap inventory fractions in Table 3 and the estimate of clad damage. The release from melted fuel should be based on the early in-vessel phase data in Table 1 and the percentage of the fuel affected.
2.
If no or minimal' fuel damage is postulated for the limiting event, the activity released should be the maximum coolant activity allowed by technical specification. The iodine concentration in the primary coolant is assumed to correspond to the following two cases in the NSSS vendor's standard technical specifications:
a.
the concentration that is the maximum value (typically 4.0 pCl/gm DE l-131) permitted and corresponds to the conditions of an assumed pre-accident spike; and b.
the concentration that is the maximum equilibrium value (typically 0.2 pCi/gm DE l-131) permitted for continued full power operation.
3.
The activity released from the fuel should be assumed to mix instantaneously and homogeneously in the reactor coolant. Noble gases should be assumed to enter the steam phase instantaneously.
Transport 4.
The assumptions related to the transport, reduction, and release of radioactive material to the environment are as follows:
a.
The main steam line isolation valves (MSIV) should be assumed to close in the maximum time allowed by technical specifications.
b.
The total mass of coolant released should be assumed to be that amount in the steam line and connecting lines at the time of the break plus the amount that passes through the valves prior to closure.
' The activity assumed in the analysis should be based on the activity associated with the projected fuel damage or the maximum technical specification values, whichever maximizes the radiological consequences. In determining dose equivalent 1 131 (DE l 131), only the radiolodine associated with normal operations or iodine spikes should be included. Activity from projected fuel damage should not be included.
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WORKING DRAFT -SUBJECT TO CHANGE c.
All of the radioactivity in the released coo! ant should be assumed to be released to the atmosphere within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> as a ground level release. No credit should be assumed for plateout, holdup, or dilution within facility buildings.
d.
The iodine release from the main steam line should be assumed to consist of 95% Csl as an aerosol,4.85% elemental, and 0.15% organic.
l WORKING DRAFT -SUBJECT TO CHANGE
i WORKING DRAFT - SUBJECT TO CHANGE Appendix E ASSUMPTIONS USED FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A PWR MAIN STEAM LINE BREAK ACCIDENT This appendix provides assumptions acceptable to the staff for evaluating the radiological consequences of a main steam line break accident at PWR light water reactors. These assumptions supplement the guidance provided in the main body of this guide.
Source Terms j
1.
Acceptable assumptions regarding core inventory and the release of radionuclides from the fuel are provided in Section C.3 of this regulatory guide. The release from breeched fuel clad should be based on the gap inventory fractions in Table 3 and the estimate of clad damage. The release from melted fuel should be based on the early in-vessel phase data in Table 2 and the percentage of the fuel affected. The fuel damage estimate should assume that the highest worth control rod is stuck at its fully withdrawn position.
2 2.
If no or minimal fuel damage is postulated for the limiting event, the activity released should be the maximum coolant activity allowed by technical specification. Two cases of iodine spiking should be assumeJ.
a.
A reactor transient has occurred prior to the postulated MSLB and has raised the primary coolant iodine concentration to the maximum value (typically 60 pCi/gm DE l-131) permitted by the technical specifications (i.e., a preaccident iodine spike case),
b.
The primary system transient associated with the MSLB causes an lodine spike in the primary system. The increase in primary coolant iodine concentration is estimated using a spiking model which assumes that the lodine release rate from the fuel rods to the primary coolant (expressed in curies per unit time) increases to a value 500 times greater than the release rate corresponding to the iodine concentration at the equilibrium value (typically 1.0 pCl/gm DE l-131) specified in technical specifications (i.e., concurrent iodine spike case). A concurrent iodine spike need not be considered if fuel damage is postulated. The assumed iodine spike duration should be 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 3.
The activity released from the fuel should be assumed to be released instantaneously and homogeneously through the primary coolant.
4.
The chemical form of radiciodine released from the fuel should be assumed to be 95% cesium lodide (Csi),4.85 percent elemental iodine, and 0.15 percent organic iodide. lodine releases via the steam j
generators should be assumed to 100% elemental. These fractions apply to iodine released due to fuel damage and to iodine released during normal operations including iodine spiking.
l
'.. Facilities licensed with, or applying for, attemate repair criteria (ARC) should use this section in conjunction with the applicable guidance in Regulatory Guide <DG1074>, Steam Generator Tube Integrity 8 The activity assumed in the analysis should be based on the activity associated with the projected fuel damage or the maximum technical specification values, whichever maximizes the radiological consequences. In determining dose equivalent 1-131 (DE l-131), only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel damage should not be included.
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WORKING DRAFT - SUBJECT TO CHANGE Transport' 5.
The assumptions related to the transport, reduction, and release of radioactive material to the environment are as follows:
a.
For facilities that have not implemented attemate repair criteria (ARC)(Ref. <DG1074>), the primary-to-secondary leak rate in the steam generator (s) should be assumed to be the leak rate limiting condition for operation specified in the technical specifications. For facilities with traditional per generator and all generators specifications, the leakage should be apportioned between affected and unaffected steam generator (s) in such a manner that the calculated dose is maximized.
b.
The density used in converting volumetric leak rates (e.g., gpm) to mass leak rates (e.g.,
Ibm /hr) should be consistent with the basis of the parameter being converted. The ARC leak rate correlations are generally based on the co!!ection of cooled liquid. Surveillance tests and facility instrumentation used to show compliance with leak rate technical specifications are typically based on cooled liquid. In most cases, the density should be i
assumed to be 1.0 gm/cc (62.4 lbm/ft3).
c.
The pdmary-to-secondsry leakage should be assumed to continue until the primary system pressure is less than the secondary system pressure, or until the temperature of the j
leakage is less than 100 *C (212 'F).
d.
All noble gas radionuclides released from the primary system are assumed to be released to the environment without reduction or mitigation.
e.
All iodine and particulate radionuclides released from the primary system via the faulted steam generator (s) should be assumed to be released to the environment with no mitigation.
f.
During periods of total submergence of the tubes in the non-faulted steam generator (s), the primary-to-secondary leakage should be assumed to mix with the bulk water in the steam generator (s). This leakage is released to the environment at a rate based on the steam mass flow rate from the steam generator (s).
h.
A partitioning coefficient for iodine of 100 should be assumed during periods of total sul"nergence of the tubes in the non-faulted steam generator (s). The retention of partice!ste mdienuclides in the steam generator (s) is limited by the moisture carryover from the steam gernrator(s). The decontamination of particulates should be assumed to be equal b-8.. Note: in this appendix, Ruptured refers to the state of the steam generator in which primary-to-secondary leakage rate has increased to a value greater than technical specifications. Fautted refers to the state of the steam generator in which the secondary side has been depressurized due to a MSLB such that protective system response (main steam line isolation, reactor trip, safety injection, etc.) has occurred. Partifioning Coefficient is defined as:
mass of Ig 997 Unit mass of liquid mass of l per unit mass of gas a
WORKING DRAFT --SUBJECT TO CHANGE
-~
WORKING DRAFT - SUBJECT TO CHANGE particulate radionuclides in the steam generator (s) is limited by the moisture carryover from l
the steam generator (s). The decontamination of particulates should be assumed to be L
equal to:
00 %
DF=
%MC l
where, %MC is the percent moisture carryover for the steam generator (s) for the thermodynamic conditions associated with the event. For example, a moisture carryover of 0.25% would correspond to a DF=400.
l i.
Operating experience and analyses have shown that for some steam generator designs, tube uncovery may occur for a short period following any reactor trip (Ref. <lN88-3D).
Primary-to-secondary leakage that occurs during these periods should be assumed to be released to the environment without mixing in the steam generator bulk water and no credit l
should be taken for lodine partititioning. The impact of emergency operating procedure restoration strategies on steam generator water level need to be considered.
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6 WORKING DRAFT - SUBJECT TO CHANGE j
WORKING DRAFT - SUBJECT TO CHANGE Appendix F ASSUMPTIONS USED FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A PWR STEAM GENERATOR TUBE RUPTURE ACCIDENT'
~~ This appendix provides assumptions acceptable to the staff for evaluating the radiological 1
consequences of a steam generator tube rupture accident at PWR light water reactors. These assump'. ions supplement the guidance provided in the main body of this guide.
Source Term 1.
Acceptable assumptions regarding core inventory and the release of radionuclides from the fuel are provided in Section C.3 of this regulatory guide. The release from breeched fuel clad should be based on the gap inventory fractions in Table 3 and the estimate of clad damage. The release from melted fuel should be based on the early in-vessel phase data in Table 2 and the percentage of the fuel affected.
8 2.
If no or minimal fuel damage is postulated for the limiting event, the activity released should be the maximum coolant activity allowed by technical specification. Two cases of iodine spiking should be assumed:
a.
A reactor transient has occurred prior to the postulated SGTR and has raised the primary coolant iodine concentration tc the maximum value (typically 60 pCl/gm DE l-131) permitted by the technical specifications (i.e., a preaccident iodine spike case).
b.
The primary system transient associated with the SGTR causes an iodine spike in the primary system. The increase in primary coolant iodine concentration is estimated using a spiking model which assumes that the iodine release rate from the fuel rods to the primary coolant (expressed in curies per unit time) increases to a value 335 times greater than the release rate corresponding to the iodine concentration at the equilibrium value (typically 1.0 pCi/gm DE l 131) specified in technical specifications (i.e., concurrent iodine spike case). A concurrent iodine spike need not be considered if fuel damage is postulated. _The assumed lodine spike duration should be 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
3.
The activity released from the fuel, if any, should be assumed to be released instantaneously and homogeneously through the primary coolant.
4.
lodine releases via the steam generators should be assumed to 100% elemental.
'.. Facilities licensed with, or applying for, attemate plugging criteria should use this section in conjunction with the applicable guidance in Regulatory Guide <DG1074>, Steam Generator Tube Integrity 2 The activity assumed in the analysis should be based on the activity associated with the projected fuel damage or the maximum technical specification values, whichever maximizes the radiological consequences. In determining dose equivalent 1-131 (DE l-131), only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel damage should not be included.
WORKING DRAFT -- SUBJECT TO CHANGE
WORKING DRAFT - SUBJECT TO CHANGE Transport * -
5.
The assumptions related to the transport, reduction, and release of radioactive material to the environment are as follows:
a.'
The primary-to-secondary leak rate in the steam generator (s) should be assumed to be the
~
leak rate limiting condition for operation specified in the technical specifications. The leakage should be apportioned between affected and unaffected steam generator (s)in such a manner that the calculated dose is maximized.
b.
The density used in converting volumetric leak rates (e.g., gpm) to mass leak rates (e.g.,
- Ibm /hr) should be consistent with the basis of surveillance tests used to show compliance with leak rate technical specifications. These tests are typically based on cool liquid.
Facility instrumentation used to determine leakage is typically located on lines containing cool liquids. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft3).
c.
The primary-to-secondary leakage should be assumed to continue until the primary system pressure is less than the secondary system pressure, or until the temperature of the leakage is less than 100 *C (212 'F).'
d.
The release of fission products from the secondary system should be evaluated with the assumption of a coincident loss of offsite prarer.
e.
All noble gas radionuclides released from the primary system are assumed to be released to the environment without reduction or mitigation.
f.
During periods of total submergence of the tubes in the ruptured steam generator, the transport model described in this section should be utilized for iodine and particulates. This model is shown in Figure F-1 arid summarized below:
A portion of the primary-to-secondary leakage will flash to vapor, based on the thermodynamic conditions in the reactor and secondary coolant.
The leakage that flashes to vapor will rise through the bulk water of the steam generator and enter the steam space. Credit may be taken for scrubbing in the generator, using the models in NUREG-0409, lodine Behaviorin a PWR Cooling System Following a Postulated Steam Generator Tube Rupture Accident.
The leakage tha' does not flash is assumed to mix with the bulk water and will t
e become vapor at a rate which is the function of the steaming rate and the partition coefficient.
I s.. Note: In this appendix, Ruptured refers to the state of the steam generator in which primary-to-secondary leakage rate has increased to a value greater than technical specifications. Partitioning Coefficient is defined as:
mass of 1, per unit mass of liquid p
mass of la per unit mass of gas WORKING DRAFT - SUBJECT TO CHANGE
WORKING DRAFT -SUBJECT TO CHANGE 11.
A partitioning coefficient for iodine of 100 may be assumed during periods of total submergence of the tubes. The retention of particulate radionuclides in the steam 9enerator(s)is limited by the moisture carryover from the steam generator (s). The decontamination of particulates should be assumed to be equal to:
DF = 100*/o
%MC whem, %MC is the percent moisture carryover for the steam generator for the thermodynamic conditions associated with the event.
Figure F-1 Transport Model Release Steam Space p
J L (Scrubbing)
Primary (Partitioning)
Leakage
} Bulk Water g.
For the non-ruptured steam generator (s) used to perform post-event cooldown, primary coolant leakage should be assumed to mix with the bulk water without flashing. A partitioning coefficient for iodine of 100 may be assumed during periods of total submergence of the tubes. The retention of particulate radionuclides in the steam generator (s) is firnited by the moisture carryover from the steam generator (s), as discussed
- above, h.
Operating experience and analyses have shown that for some steam generator designs, tube uncovery may occur for a short period following any reactor trip (Ref. <lN88-3k).
Primary-to secondary leakage that occurs during these periods should be assumed to be released to the environment without mixing in the steam generator bulk water and no credit should be taken for iodine partititioning. The impact of emergency operating procedure restoration strategies on steam generator water level need to be considered.
i WORKING DRAFT --- SUBJECT TO CHANGE
WORKING DRAFT - SUBJECT TO CHANGE Appendix G ASSUMPTIONS USED FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A PWR LOCKED ROTOR ACCIDENT'
~ ~ '
This appendix providss assumptions acceptable to the staff for evaluating the radiological consequences of a locked rotor accident at PWR light water reactors. These assumptions supplement the l
guidance provided in the main body of this guide.
Source Term 1.
Acceptable assumptions regarding core inventory and the release of radionuclides from the fuel are provided in Section C.3 of this regulatory guide. The release from breeched fuel clad should be based on the gap inventory fractions in Table 3 and the estimate of clad damage. The release from melted fuel should be based on the early in-vessel phase data in Table 2 and the percentage of the fuel affected.
l 2.
If no fuel damage is postulated for the limiting event, a radiological analysis is not required as the consequences of this event are bounded by the consequences projected for the main steam line break and steam generator tube rupture.
3.
Jhe activity released from the fuel should be assumed to be released instantaneously and homogeneously through the primary coolant.
i 4.
The chemical form of radiolodine released from the fuel should be assumed to be 95% cesium iodide (Csi),4.85 percent elementaliodine, and 0.15 percent organic iodide. lodine releases via the steam generators should be assumed to 100% elemental. These fractions apply to iodine released due to fuel damage and to iodine released during normal operations including lodine spiking.
Release Transport 5.
The assumptions related to the transport, reduction, and release of radioactive material to the environment are as follows:
a.
The primary-to-secondary leak rate in the steam generators should be assumed to be the leak rate limiting condition for operation specified in the technical specifications. The leakage should be apportioned between the steam generators in such a manner that the calculated dose is maximized.
b.
The density used in converting volumetric leak rates (e.g., gpm) to mass leak rates (e.g.,
Ibm /hr) should be consistent with the basis of surveillance tests used to show compliance with leak rate technical specifications. These tests are typically based on cool liquid.
Facility instrumentation used to determine leakage is typically located on lines containing cool liquids. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft3).
c.
The primary-to-secondary leakage should be assumed to continue until the primary system pressure is less than the secondary system pressure, or until the temperature of the
'.. Facilities licensed with, or applying for, alternate plugging criteria should use this section in conjunction with the applicable guidance in Regulatory Guide <DG1074>, Steam Generator Tube integrity WORKING DRAFT --- SUBJECT TO CHANGE
WORKING DRAFT - SUBJECT TO CHANGE leakage is less than 100 *C (212 *F).
1 I
d.
The release of fission products from the secondary system should be evaluated with the assumption of a coincident loss of offsite power.
~
All noble gas radionuclides released from the primary system are assumed to be released e.
to the environment without reduction or mitigation, f.
During periods of total submergence of the tubes steam generators, the primary-to-secondary leakage should be assumed to mix with the bulk water in the steam generators.
This leakage is released to the environment at a rate based on the steam mass flow rate from the steam generators.
2 g.
A partitioning coefficient of 100 should be assumed for elementallodine during periods of total submergence of the tubes. The retention of particulate radionuclides in the steam generators is limited by the moisture carryover from the steam generators. The decontamination of particulates should be assumed to be equal to:
DF = 100%
%MC where, %MC is the percent moisture carryover for the steam generator (s) for the thermodynamic conditions associated with the event. For example, a moisture carryover of 0.25% would correspond to a DF=400.
i h.
Operating experience and analyses have shown that for some steam generator designs, tube uncovery may occur for a short period following any reactor trip (Ref. <!N88-31>).
Primary-to-secondary leakage that occurs during these periods should be assumed to be released to the environment without mixing in the steam generator bulk water and no credit should be taken for iodine partititioning. The impact of emergency operating procedure restoration strategies on steam generator water level need to be considered.
- Partitioning Coetticientis defined as:
. P.C.= "*** *'f la per unit mass of gas mass o WORKING DRAFT - SUBJECT TO CHANGE
3 i
WORKING DRAFT - SUBJECT TO CHANGE Appendix H ASSUMPTIONS USED FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A PWR ROD EJECTION ACCIDENT' This appendix provides assumptions acceptable to the staff for evaluating the radiological consequences of a rod ejection accident at PWR light water reactors. These assumptZns supplement the guidance provided in the main body of this guide.
l Source Term 1.
Acceptable assumptions regarding core inventory and the release of radionuclides from the fuel are provided in Section C.3 of this regulatory guide. The release from breeched fuel clad should be based on the gap inventory fractions in Table 3 and the estimate of clad damage. The release from melted fuel should be based on the early in-vessel phase data in Table 2 and the percentage of the fuel affected.
2.
If no fuel damage is postulated for the limiting event, a radiological analysis is not required as the consequences of this event are bounded by the consequences projected for the LOCA, main steam line break and steam generator tube rupture.
3.
Two release cases are to be considered. In the first,100% of the activity released from the fuel should be assumed to be released instantaneously and homogeneously through the containment atmosphere. In the second,100% of the activity released from the fuel should be assumed to be completely dissolved in the primary coolant an available for release to the secondary system.
4.
The chemical form of radiolodine released to the containment atmosphere should be assumed to be 95% cesium iodide (Csi),4.85 percent elemental iodine, and 0.15 percent organic iodide. If containment sprays do not actuate or are terminated prior to accumulating sump water, or if the containment sump pH is not controlled at values of 7 or greater, the iodine species should be evaluated on an individual case basis. Evaluations of pH should consider the effect of acids created during the LOCA event, e.g., pyrolysis and radiolysis products. With the exception of elemental and organic iodine and noble gases, fission products should be assumed to be in particulate form.
5.
lodine releases via the steam generators should be assumed to 100% elemental.
Transport From Containment 6.
The assumptions related to the transport, reduction, and release of radioactive material in I-J (rom the containment are as follows:
a.
Reduction in the amount of radioactive material available for leakage from the containment j
due to natural deposition, containment sprays, recircult fog filter systems, dual containments, or other engineered safety features trat be taken into account. Refer to Appendix A (LOCA) in this guide for nuidance on acceptable methods and assumptions.
'..iacilities licensed with, or applying for, alternate plugging criteria should use this section in conjunction with the applicable guidance in Regulatory Guide <DG1074>, Steam Generator Tube Integrity WORKING DRAFT - SUBJECT TO CHANGE
-~.-.
. -..~
WORKING DRAFT - SUBJECT TO CHANGE b.
The containment should be assumed to leak at the leak rate incorporated in the technical specifications at peak accident pressure for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and at 50% of this leak rate for the remaining duration of the accident. Peak accident pressure is the maximum pressure defined in the technical specifications for containment leak testing. Leakage from subatmospheric containments is assumed to be terminated when the containment is brought to a subatmospheric condition as defined in technical specifications.
Transport From Secondary System 7.
The assumptions related to the transport, reduction, and release of radioactive materialin and from the secondary system are as follows:
A leak rate equivalent to the primary-to-secondary leak rate limiting condition for operation a.
specifitK1 in the technical specifications should be assumed to exist until the primary system pressure remains less than the secondary system pressure, b.
The density used in converting volumetric leak rates (e.g., gpm) to mass leak rates (e.g.,
Ibm /hr) should be consistent with the basis of surveillance tests used to show compliance with leak rate technical specifications. These tests are typically based on cooled liquid.
Facility instrumentation used to determine leakage is typically located on lines containing cool liquids. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft3).
c.
All noble gas radionuclides released from the primary system are assumed to be released to the environment without reduction or mitigation.
d.
During periods of total submergence of the tubes steam generators, the primary-to-secondary leakage should be assumed to mix with the bulk water in the steam generators.
This leakage is released to the environment at a rate based on the steam mass flow rate from the steam generators, e.
A partitioning coefficient' of 100 should be assumed for elemental iodine during periods of total submergence of the tubes. The retention of particulate radionuclides in the steam generators is limited by the moisture carryover from the steam generators. The decontamination of particulates should be assumed to be equal to:
DF=
%MC where, %MC is the percent moisture carryover for the steam generator (s) for the thermodynamic conditions associated with the event. For example, a moisture carryover of 0.25% would correspond to a DF=400, f.
Operating experience and analyses have shown that for some steam generator designs, tube uncovery may occur for a short period following any reactor trip (Ref. <lN88-31>).
I-
- Partitioning CoeMicientis defined as:
P.C.= ***** # " U *****'
p mass of la per unit mass of gas WORKING DRAFT - SUBJECT TO CHANGE
WORKING DRAFT -SUBJECT TO CHANGE
~
Primary-to-secondary leakage that occurs during these periods should be assumed to be released to the environment without mixing in the steam generator bulk water and no credit should be taken for lodine partititioning. The impact of emergency operating procedure restoration strategies on steam generator water level need to be considered.
f WORKING DRAFT -SUBJECT TO CHANGE
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