POLICY ISSUE 

December 19, 1994 (Information) SECY-94-302 

HEO: The Commissioners 

FRMf: James M. Taylor 

Executive Director for Operations 

-SU.BJEC: SOURCE TERM-RELATED TECHNICAL AND LICENSING 

ISSUES PERTAINING TO EVOLUTIONARY AND PASSIVE 

LIGHT-WATER-REACTOR DESIGNS 

PURPOSE: 

To inform the Commission of staff positions pertaining to source term-related technical and licensing issues and the staff's methodologies for using these staff positions in the evolutionary and passive light-water-reactor (LWR) design certification reviews. 

BACKGROUND: 

In SECY-90-016, "Evolutionary Light-Water-Reactor (LWR) Certification Issues and Their Relationship to Current Regulatory Requirements," the staff made a recommendation to the Commission on the use of the physically-based source 

term. For evolutionary LWRs, the staff proposed to consider any deviations from current methodology for calculating 10 CFR Part 100 doses on a case-by case basis using engineering judgment and updated information on source term and equipment reliability. The Commission approved this recommendation in a staff requirements memorandum dated June 26, 1990. 

Contacts: Jay Y. Lee, NRR 

504-1080 

John A. Kudrick, NRR 

504-2871 

SECY NOTE: TO BE MADE PUBLICLY AVAILABLE IN 10 WORKING DAYS FROM THE DATE 

OF THIS PAPER.
The Commissioners

In SECY-92-127, 'Revised Accident Source Terms for Light-Water Nuclear Power 

Plants,' the staff provided the Commission a draft report (later published as 

draft NUREG-1465) describing revised accident source terms for light-water 

nuclear power plants to replace those of Technical Information Document (TID) 

-14844 which was published in 1962. The staff also described the process it 

would use to finalize the report, including both a public comment and peer 

review period. Any specific applications of the new source term prior to the 

completion of this process were to be submitted to the Commission for 

preliminary review. The staff expects to issue the final version of NUREG-1465 

in late 1994. 

In this paper, the staff notes that the review process for NUREG-1465 is 

essentially complete and presents its positions on (1) the closure of source 

term-related issues in its SERs for the EPRI requirements documents for both 

evolutionary and passive plant designs and (2) the generic implementation of 

source term-related issues in evolutionary and passive LWR design 

certification reviews. The staff is not requesting a review by the Commission 

of these positions because they are primarily technical applications of 

previous Commission policy decisions. The application of the revised source 

terms is specific to the design of the reactor. Therefore, these final staff 

positions are those taken in the staff's review of the Asea Brown Boveri

Combustion Engineering (ABB-CE) System 80+ and are those that will be taken 

for the passive LWR design reviews. 

The final NUREG-1465 will not alter the staff's conclusion in the System 80+ 

FSER in regard to its acceptance of the System 80+ design for meeting the dose 

reference values given in 10 CFR Part 100 and the control room operator dose 

limit specified in GDC 19. 

DISCUSSION: 

The current reactor accident source terms for fission-product release from the 

reactor core into the containment are set forth in Regulatory Guide 1.3, 

"Assumptions Used for Evaluating the Potential Radiological Consequences of a 

Loss of Coolant Accident for Boiling Water Reactors," and Regulatory Guide 

1.4, "Assumptions Used for Evaluating the Potential Radiological Consequences 

of a Loss of Coolant Accident for Pressurized Water Reactors." These source 

terms were derived from TID-14844. The regulatory guide source terms are used 

in conjunction with postulated design-basis accidents (DBAs). Revised reactor 

accident source terms have been proposed in draft NUREG-1465, "Accident Source 

Terms for Light-Water Nuclear Power Plants." These source terms were derived 

from an examination of a set of accident sequences for current LWR designs and 

reflect the current understanding of severe accidents and fission-product 

behavior. 

The staff intends to use the reactor accident source terms given in NUREG-1465 

in radiological consequence assessments in the following areas of evolutionary 

and passive LWR design certification reviews:

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(1) equipment qualification 

(2) control room habitability 

(3) engineered safety features atmosphere cleanup systems (4) primary containment leak rate 

(5) containment isolation timing 

(6) post-accident sampling 

(7) shielding and vital area access 

In this paper the staff identifies its significant technical positions relative to the implementation of the revised source terms for evolutionary 

and passive LWR designs. These positions are more fully discussed in 

Attachment 1. On the basis of these positions, the staff closed out all 

source term-related open issues in its SER for the EPRI Requirements Document 

for passive plant designs and the ABB-CE System 80+ design. These positions will provide the bases for the staff's review of passive LWR certification 

applications. 

The revised source terms are not proposed as a requirement for existing 

plants. Preliminary indications are that the revised release estimates, in 

conjunction with improved insights on timing and best estimate credit for 

fission product removal mechanisms, would not result in more stringent 

requirements if applied to existing plants, and may support relaxation of 

certain present regulatory requirements, particularly those associated with 

time of fission-product appearance in the reactor containment. 

Because it reflects our current understanding of source term behavior, the 

staff plans to make the revised source term information available to existing 

licensees for voluntary proposals to modify current requirements that may be 

overly conservative. However, due to the large number of design and 

operational areas impacted by the source term and the plant specific nature of 

these impacts, the staff has not determined at this time the extent to which 

these insights could be implemented for existing plants. A memorandum to the 

Commission entitled "Use of NUREG-1465 Source Term at Operating Reactors," 

dated September 6, 1994, discussed the staff's plans for evaluating the 

application of the revised source term to operating reactors, including 

interaction with the public and industry. 

In Attachment 1, the staff discusses 12 source term-related issues of both a 

technical and licensing nature that pertain to either evolutionary LWRs, 

passive LWRs, or both. The staff previously identified all of these issues in 

SECY-93-087, "Policy, Technical, and Licensing Issues Pertaining to 

Evolutionary and Advanced Light-Water (ALWR) Designs," dated April 2, 1993. 

Of these 12, the staff highlights (in italics) the 5 most significant issues 

(Issues I through 5) in the following summary discussion.

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Draft NUREG-1465. "Accident Source Terms for Light-Water Nuclear Power Plants" 

In SECY-92-127, the staff provided, for the Commission's Information, a draft 

staff report containing proposed revised accident source terms for light-water 

nuclear power plants to replace those of Regulatory Guides 1.3 and 1.4. The 

draft report was issued for public comment in July 1992 as NUREG-1465. Draft 

NUREG-1465 addressed the current understanding of LWR severe accident and 

fission-product behavior, including the quantity and chemical forms of fission 

products released into the reactor containment following a severe reactor 

accident as well as fission-product release timing. Since issuing draft 

NUREG-1465, the staff has received more than 200 comments from 20 U.S. and 

foreign organizations. 

The most significant public comments concerned the release fractions of the 

less volatile and non-volatile elements such as barium, strontium, and cerium 

during the early in-vessel release phase. The release fractions in draft 

NUREG-1465 of the less volatile and non-volatile elements are about 10 to 100 

times higher than those proposed by EPRI in its Utility Requirements Document 

for Advanced Light-Water Reactors. In response to comments, the staff re

examined the release of non-volatile elements during the early in-vessel 

phase, and determined that the "mean" values of draft NUREG-1465 were not 

appropriate since they were well in excess of other measures of the 

uncertainty distribution, such as the 75th percentile. 

Final NUREG-1465 is expected to utilize the 75th percentile values to account 

for the considerable uncertainty regarding the releases of the non-volatile 

elements in the early in-vessel phase. This will significantly reduce the 

release fractions from those specified in draft NUREG-1465 for the non

volatile elements in the early in-vessel phase. 

The staff found in its radiological consequence assessments that the less 

volatile and non-volatile elements play a less significant role, contributing 

less than a few percent of the overall radiological consequences resulting from a DBA, compared to the other major nuclides such as noble gases, iodine, 

and cesium. Therefore, the staff believes that revisions to the release 

fractions of the less volatile and non-volatile elements in final NUREG-1465 

will not materially change the ongoing staff reviews of the passive LWR 

designs. 

Draft NUREG-1465 indicated that iodine is released initially into containment 

primarily (95%) in particulate form as CsI, with about 5% in volatile form, as 

elemental iodine (12). If the pH of the water in the containment is 

maintained at a value of 7 or above, elemental iodine would remain at a low 

level. Organic forms of iodine, primarily methyl iodide, are formed largely 

from reactions of organic compounds with elemental iodine, and may contribute 

significantly to offsite doses because they are not easily removed by 

engineered safety features such as sprays or filters. 

Based on past research, about 3 percent of the elemental iodine present is 

expected to be converted to organic iodine. As discussed in the DBA 

assessment section below, the staff intends to assume, where the pH of the

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water in the containment is controlled, that 0.15 percent (3 percent of 5 percent) of the core iodine inventory is in organic form and available for leakage from the containment to the environment following a DBA. Although draft NUREG-1465 did not address the formation of organic iodine, final NUREG 1465 will address this issue. 

The fission-product release timing differs with reactor type and the bounding design basis accident sequence chosen for the source term application. The staff will review each ALWR design on a design specific basis to determine appropriate fission-product release timing. 

Most of the remaining public comments are largely favorable regarding draft NUREG-1465. The staff briefed the Commission on the status of the proposed source term update, including the discussion of major comments received on the draft NUREG-1465, on August 3, 1993. The staff also discussed it with ACRS on September 9, 1994. The staff expects to issue the final version of NUREG-1465 

in late 1994. 

Draft NUREG-1465 also contained a section on mechanisms for removing fission products from the reactor containment atmosphere following a severe reactor accident that was based on the realistic estimates of the natural processes as used in the NUREG-1150 ("Severe Accident Risks: An Assessment for 5 U.S. Nuclear Power Plants") analyses. To obtain more comprehensive information on removal mechanisms to be used in the licensing reviews, the staff is continuing to explore this area with NRC contractor assistance. 

The staff's positions on the source term issues discussed in this paper are based on NRC contractor findings in NUREG/CR reports, preliminary findings in NRC contractor reports, extensive consultation with these contractors, the staff's engineering judgments, and the need to maintain an appropriate margin of safety. 

Design-Bas is-Accident Assessment 

To evaluate the passive LWR submittals, the staff will utilize the current insights from source term research as described in draft NUREG-1465 regarding fission-product releases into the containment. In determining the effects of such removal mechanisms as sprays, filters, plateout, and aerosol deposition, the staff will use the best available information, including engineering judgment, for the applicable parameters. 

The use of physically-based source terms as described in draft NUREG-1465 constitutes a major departure from the use of the current regulatory guide source terms. The proposed source terms lead to a qualitatively different fission-product distribution, which is a result of (1) physically-based composition, release timing, and rate of releases and (2) inclusion of removal mechanisms based on natural processes. 

In Attachment 2, the staff presents the revised source terms to be used in conjunction with DBAs for reviewing passive LWR designs, including their

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release timing as well as duration for boiling-water reactors (BWRs) (Table 1) 

and for pressurized-water reactors (PWRs) (Table 2). The information in these 

tables, taken from the proposed final NUREG-1465, may be revised when NUREG

1465 is issued in final form. The staff does not expect that any revision 

will materially change its evaluation or conclusions pertaining to these 

designs. 

The staff-proposed source term release fractions (magnitudes) in proposed 

final NUREG-1465 are compared to those proposed in the evolutionary and 

passive plant designs in Table 3 of Attachment 2. The staff has used only the 

gap release and the early In-vessel release for design-basis calculations in 

ABB-CE System 80+ design review and will be using them for the passive LWR 

designs reviews. These source terms encompass a broad range of accident 

scenarios, including significant levels of core damage with the core remaining 

in the vessel. These would be the most severe scenarios from which the plant 

could be expected to return to a safe shutdown condition. Table 3 in 

Attachment 2 shows that the proposed final NUREG-1465 release fractions 

associated with these releases are generally comparable to those of TID-14844. 

The staff considers the inclusion of the ex-vessel and late in-vessel source 

terms to be unduly conservative for DBA purposes. Such releases would only 

result from core damage accidents with vessel failure and core-concrete 

interactions. For evolutionary and passive LWRs, the estimated frequencies of 

such scenarios are low enough that they need not be considered credible for 

the purpose of meeting 10 CFR Part 100. 

On the basis of a comparison of the gap and in-vessel releases (excluding ex

vessel releases), the staff finds that the quantities of noble gases and 

radioiodines are in reasonable agreement with those in TID-14844, while the 

proposed new source terms include substantially more of the low volatility 

solids. For the regulatory guide source terms, the staff assumed that the 

fraction of these solids released represents less than 1 percent of the 

activity available for release, while the revised source terms show 

significantly larger release fractions for these materials. The important 

difference is that the proposed new source terms are released over a period of 

1.8 (PWR) to 2.0 hours (BWR), rather than instantaneously. 

In radiological assessments of DBAs for passive light-water reactor designs, 

the staff will: 

Selectively use the source terms given in draft NUREG-1465 using only "Gap 

Release" and "Early In-Vessel Release" (excluding "Ex-Vessel Release" and 

"Late In-Vessel Release" associated with vessel failure and core-concrete 

interaction) in evaluating radiological consequences for DBAs, the DBA 

radiation environmental qualification of electrical and mechanical equipment 

important to safety, post-accident shielding, and Three Wile Island-related 

requirements (Issue 1). 

Use the chemical forms of iodine of at least 95 percent cesium iodine as 

stated in draft NUREG-1465 with 4.85 percent of elemental iodine and hydrogen 

iodide and 0.15 percent organic iodide (Issue 2).

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Severe Accident Assessment 

The staff recognizes that the new source terms have implications for the staff 

consideration of severe accidents as well. The staff's resolution of severe 

accident issues is based on the inclusion of specific design features to 

provide assurance of containment integrity for approximately 24 hours 

following the onset of core damage. Part of the basis for the 24-hour 

criterion is the fact that natural fission-product removal mechanisms can 

significantly reduce the suspended aerosol source terms present in the 

containment within this time period. 

Another aspect of the resolution of severe accident issues is the requirement 

that reasonable assurance exist that equipment needed for severe accident 

mitigation and post-accident sampling will survive in the severe accident 

environment. This was an issue raised to the Commission in SECY-90-016 and 

SECY-93-087 and approved in staff requirements memoranda dated June 26, 1990, 

and July 21, 1993. It is the responsibility of the design certification 

applicant to demonstrate that all such equipment can survive the radiation 

environment following a severe accident. As discussed in Enclosure 1, the 

staff believes that the design-basis source terms described above may not 

represent reasonable surrogates for the environmental conditions (such as 

pressure, temperature, radiation, and humidity) in a severe accident. For a 

severe accident, the design certification applicant should include the 

contribution of fission products that may be released ex-vessel and late in

vessel. For those reactors that have a reliable ex-vessel flooding mechanism 

for the core debris, this contribution may be significantly reduced by aerosol 

scrubbing. 

In radiological assessments of equipment survivability as a result of a severe 

reactor accident, the staff will require that the equipment and features 

needed for severe accident prevention, mitigation, and post-accident sampling 

be designed to provide a reasonable level of confidence that they will operate 

in a severe accident environment. This environment would include the ex

vessel release, with proper credit for design features to mitigate that 

release, and the late in-vessel release In addition to the releases for a DBA 

(Issue 3). 

Finally, the probabilistic safety assessment (PSA) for each advanced reactor 

includes an evaluation of the dose consequences associated with a variety of 

core melt scenarios and containment failure modes. These calculations were 

performed with the current understanding of source term behavior. While these 

assessments generally contain some discussion of uncertainties in fission

product behavior, they do not include any explicit provision for conservatism. 

As is generally the case In PRAs, these calculatiDns are done on a realistic 

basis. 

Design-SRecific Reviews and Staff Positions 

The staff has completed its review of the General Electric (GE) advanced 

boiling water reactor (ABWR) and the ABB-CE System 80+ designs and is 

currently reviewing the Westinghouse AP600 reactor design.

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ABB-CE initially proposed to use the TID-14844 source terms. However, it 

subsequently submitted to the staff an amendment to its standard safety analysis report (SSAR) incorporating the revised accident source terms as 

described in draft NUREG-1465. 

The System 80+ design provided the same mitigating features for the 

radiological consequences resulting from a DBA as those provided for the 

currently operating PWRs (containment spray, engineered safety features (ESF) filtration, etc.), except that safety-grade charcoal adsorbers are not 

provided. ABB-CE demonstrated in its SSAR that the System 80+ design will 

meet the dose reference values given in 10 CFR Part 100 and GDC 19 using the 

source terms provided in draft NUREG-1465 in conjunction with the use of 

atmospheric relative concentrations (x/Qs) provided in its Tier I Design Document. The staff reviewed the System 80+ design using the staff's 

positions discussed in this Commission paper and concluded that the design will meet the above relevant dose reference values. 

The final NUREG-1465 will not alter the staff's conclusion in the System 80+ 

FSER in regard to its acceptance of the System 80+ design for meeting the dose 

reference values given in 10 CFR Part 100 and the control room operator dose 

limit specified in GDC 19. As shown in Table 3 of Attachment 2, the fission

product release fractions given in final NUREG-1465 are the same (for noble 

gases, iodine, and cesium) or lower than those given in draft NUREG-1465 (for non-volatile elements). For fission-product chemical forms, the percent of 

core iodine inventory in organic form is reduced to 0.15 in final NUREG-1465 

from 0.25 percent used in the CE System 80+ design review. There were no 

changes made in fission-product release timing for PWR in final NUREG-1465 

from that in draft NUREG-1465. 

For the AP600 design, Westinghouse proposed the same accident source terms as 

those proposed by EPRI for passive plants. However, the staff will use the 

revised source terms (fission-product release fractions and chemical forms) as 

described in final NUREG-1465 for the AP600 design review. The staff is 

currently evaluating the fission-product release timing proposed by Westinghouse. The staff realizes that fission-product release timing is 

dependent upon reactor design and reactor design basis accident sequences chosen for the source term application. The fission-product release timing given in final NUREG-1465 is based on the current LWR designs. This issue is 

discussed in more detail in Item 6 of Attachment 1. 

Unlike the current generation of operating PWRs, the AP600 design does not 

include either an ESF filtration (e.g., charcoal adsorbers) or a containment 

spray system. In SECY-93-087 (Item III.F, Radionuclide Attenuation), the 

staff informed the Commission that it is still evaluating the need for a 

containment spray system for the passive plant designs. The staff will 

review the AP600 radiological consequence assessment to determine if the dose 

reference values in 10 CFR Part 100 and the control room operator dose 

criteria of General Design Criterion 19 of Appendix A to 10 CFR Part 50 can be 

met without a containment spray system. That determination will involve 

evaluation of aerosol removal rates by natural deposition mechanisms,

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containment leak rates, and post-accident pH control of the water in the 

containment. This is addressed in more detail as Issue 9 in Attachment 1. 

NUREG-1242, "NRC Review of Electric Power Research Institute's Advanced Light 

Water Reactor Utility Requirements Document," resolves the major source term 

related issues in principle except iodine chemical species for BWR, amount of 

gap activity, magnitude of low- and non-volatile fission products released, 

and fission-product release timing. As discussed in NUREG-1242, it was 

expected that these differences would be resolved by the issuance of final 

NUREG-1465. The numbers presented in Table 3 of Attachment 2 show that the 

positions being proposed for final NUREG-1465 is much closer to the EPRI 

positions. Therefore, the staff believes that issuance of final NUREG-1465 

will resolve these differences. 

GE demonstrated in its SSAR that the ABWR design will meet the offsite dose 

reference values given in 10 CFR Part 100 using the current TID-14844 source 

terms in conjunction with the use of atmospheric relative concentrations 

(X/Qs) provided in its Tier 1 Design Document. The staff reviewed the ABWR 

design, performed an independent analysis of the radiological consequences 

resulting from a postulated DBA, and concluded in the FSER that the ABWR 

design will meet the dose reference values given in 10 CFR Part 100. 

In its review, the staff accepted the ABWR design without a main steamline 

leakage control system (LCS) that is designed to process main steamline 

leakage through main steamline isolation valves following a DBA. The staff 

also allowed credit for removal of radioactive iodine in main steamline 

leakage by holdup and plateout in the main steamline and condenser following a 

DBA. 

As discussed in Attachment 1, these departures from the guidance of Regulatory 

Guide 1.96, "Design of Main Steam Isolation Valve Leakage Control Systems for 

Boiling Water Reactor Nuclear Power Plants," and the principle of not allowing 

credit for non-seismically qualified systems are based on the conclusion that 

these systems are expected to retain sufficient structural integrity to remain 

effective for fission-product holdup and plateout throughout a DBA. 

The staff will accept the passive BWR plant design without a LCS and allow an 

appropriate credit for iodine removal In the main steamlfne and condenser 

following a DBA (Issue 4). 

In its SBWR design, GE initially proposed the same accident source terms as 

those published in draft NUREG-1465, but amended its SSAR to reflect the EPRI 

accident source terms. The SBWR design provides non-safety-grade containment 

spray systems without an ESF filtration system. This reduction of safety 

margin for mitigating radiological consequences may be compensated for by the 

secondary containment ("safety envelope") design. The safety envelope is a 

reinforced-concrete structure within the reactor building that forms an 

envelope completely surrounding the primary containment. It is designed to be 

capable of periodic testing to ensure its intended performance. The design 

also relies on lower core power density (lower reactor power) and slower, 

delayed releases of fission products into the containment.

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The staff is not reviewing the source term related technical and licensing issues for the SBWR design at the present time. However, when the staff 

resumes its review of the SBWR design, the staff will allow appropriate credit for the SBWR safety envelope based on fission-product holdup and decay within 

this envelope if (1) the vendor specifies that the secondary containment 

leakage and mixing performance be consistent with the values used by the staff in its radiological assessment and (2) the COL combined license applicant incorporates the secondary containment leakage value specified by the vendor 

into the plant-specific technical specifications (Issue 5). 

The staff has completed its review of source term-related issues in the EPRI requirements documents for evolutionary and passive plant designs and in the 

System 80+ design. It is currently reviewing the AP600 design. The staff 

positions presented in this paper will form a basis for closing all 

source term-related open issues associated with the passive ALWR designs. The 

staff does not expect that the publication of NUREG-1465 in final form will 

materially change its evaluation or conclusions pertaining to these designs. 

Margins of Conservatism 

The use of the revised source term information is an important departure from 

previous practice. The new approach will employ a physically-based source 

term based on substantial research and experience gained over two decades. 

The TID-14844 non-mechanistic methodology with the staff's application of 

conservative assumptions was intended to ensure that future plants would 

provide sufficient safety margins even with the recognized uncertainties 

associated with accident sequences and equipment reliability. While the TID

14844 source term has served its intended purpose, research started prior to the Three Mile Island Accident provides substantial information on plant behavior under accident conditions. The results of this research are 

incorporated in draft NUREG-1465. 

The staff believes that accident research insights and information provided in 

draft NUREG-1465 and this paper provide a sound, contemporary basis for 

reactor accident mitigation system designs. While the new information may lead to relaxation in some aspects of the design, it also provides safety benefits by removing unrealistically stringent testing requirements. For 

example, the TID-14844 source term specifies the instantaneous release of 

fission-products into the primary containment while reactor accident research 

insights indicate that releases occur over a period of hours. This change, when used as a partial basis for less stringent closure requirements for fast

acting isolation valves, could result in enhanced reliability and integrity of 

these valves. Another example includes giving credit for fission-product plateout in the ABWR main steamlines and condenser when no credit has 

previously been given. This is based on our improved understanding of iodine 

transport mechanisms and iodine behavior from the revised accident source term 

research insights. The credit for fission-product plateout in the main 

steamlines and condenser could also result in enhanced reliability and 

integrity of the main steam isolation valves, by allowing the staff to use the 

higher leak rates and less stringent closure requirements.

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The Commissioners

The NUREG-1465 accident research insights strongly link the revised iodine 

chemical forms discussed in this paper with the ability to control the pH of 

the water in the containment above 7.0 during the course of an accident. If 

the proposed plant design features provide for adequate pH control based on a 

consideration of the various sources of acid additions that could occur during 

the course of an accident, the staff will not need to consider re-evolution of 

elemental iodine. Accordingly, there may not be a need to consider accident 

mitigation features such as containment sprays and charcoal adsorbers. If, 

however, a particular design does not provide adequate pH control, the staff 

will require the addition of non-safety grade charcoal adsorbers or a 

containment spray system. 

With considerations such as those described above, the staff believes that the 

use of the new source terms provides a reasonable basis for establishing 

acceptable designs. 

SUMMARY: 

The methodology for implementing the revised source terms gives credit for 

improved understanding of fission product release timing resulting from 

accident research as well as more realistic estimates of reductions in the 

magnitude of potential fission-product releases from containment. These 

changes will result in reduced requirements for mitigation systems. The staff 

recognizes this effect and believes that it is justified on the basis of the 

improved understanding of fission-product behavior. The revised source terms 

should provide the staff with adequate technical bases for ensuring that 

safety margin is maintained. 

The staff presents its positions on 12 source term-related issues in 

Attachment 1. These positions provided the basis for the staff to close out 

all source term related open issues in the staff's SERs for the EPRI 

Requirements Document for passive plant designs and for the ABB-CE System 80+ 

design and to proceed with its review of passive LWR plant designs. The final 

NUREG-1465 will not alter the staff's conclusion in the ABB-CE System 80+ SER.

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COORDINATION: 

The Office of the General Counsel has reviewed this paper and has no legal 

objection. The ACRS Full Committee was briefed on February 10, and March 10, 

1994. The ACRS provided its comments in a letter to the Chairman dated 

March 15, 1994. The staff has considered the ACRS comments in developing the 

staff positions contained in this paper. 

CONCLUSIONS: 

The staff believes that the revised source term as given in NUREG-1465 is 

appropriate for use in the licensing review of evolutionary and passive LWR 

designs. The staff intends to continue application of its methodology and the 

staff positions described in this position paper to the on-going ALWR 

certification reviews. The staff intends to place this information paper into 

the PDR. 

es M. aylor 

ecutive Director 

for Operations 

Attachments: 

1. Source Term-Related Technical, and 

Licensing Issues Pertaining to Evolutionary 

and Advanced Light-Water-Reactor Designs 

2. Proposed Reactor Accident Source Terms 

DISTRIBUTION: 

Commissioners 

OGC 

OCAA 

OIG 

OPA 

OCA 

REGIONS 

EDO 

ACRS 

-SECY

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Attachment 1

SOURCE TERN-RELATED POLICY, TECHNICAL. 

AND LICENSING ISSUES PERTAINING TO EVOLUTIONARY 

AND PASSIVE LIGHT-WATER-REACTOR DESIGNS 

ISSUE 1: Selective Use of Accident Source Terms Given in Draft NUREG-1465 

(for System 80+, AP600, and SBWR). 

In SECY-90-016, the staff recommended to the Commission that it approve staff 

consideration of deviations from current methodology used to calculate 10 CFR 

Part 100 doses on a case-by-case basis using engineering judgment and updated 

information on source terms and equipment reliability for the evolutionary 

plant designs. The Commission approved this recommendation in a staff 

requirements memorandum of June 26, 1990. This section of the Commission 

paper describes such deviations to selectively use the fission-product release 

fractions described in NUREG-1465 for radiological consequence assessments of 

design-basis accidents for the evolutionary and passive light-water-reactor 

designs. 

Present regulations (10 CFR Part 100) require that a fission-product release 

from the reactor core into the containment as a result of a reactor accident 

be postulated and that its radiological consequences at the exclusion area 

boundary and low population zone outer radius be evaluated assuming the 

"expected demonstrable leak rate" from the containment and the meteorological 

conditions pertinent to the reactor site, with the implicit assumption that 

the containment remains intact against the maximum credible accident. 

Footnote 1 to 10 CFR Part 100 states that the fission-product release to be 

assumed for Part 100 dose calculations should be "based on a major 

accident.. .that would result in potential hazards not exceeded by those from 

any accident considered credible. Such accidents have generally been assumed 

to result in substantial meltdown of the core with subsequent release of 

appreciable quantities of fission products." 

The current source terms are specified in Regulatory Guide 1.3, "Assumptions 

Used for Evaluating the Potential Radiological Consequences of a Loss of 

Coolant Accident for Boiling Water Reactors," and Regulatory Guide 1.4, 

"Assumptions Used for Evaluating the Potential Radiological Consequences of a 

Loss of Coolant for Pressurized Water Reactors." The source terms were 

derived from Technical Information Document (TID)-14844, "Calculation of 

Distance Factors for Power and Test Reactor Sites." Although the staff 

considered the consequences of fission-product release into the containment to 

represent the bounding reactor accident in terms of credibility, it also 

evaluated other accident types to verify that these did not result in greater 

consequences. This practice led the staff eventually to develop and consider 

a group of accidents, referred to as "design-basis accidents (DBAs.)." 

The DBAs routinely evaluated by the staff as part of the license review 

include (1) loss-of-coolant accident (LOCA), (2) fuel handling accident, (3) 

steam generator tube rupture accident (PWR), (4) main steamline break outside
2 

the containment, (5) control rod drop accident (BWR), (6) control rod ejection 

accident (PWR), (7) failure of small lines carrying primary coolant outside 

the containment, and (8) spent fuel cask drop accident. The staff used the 

calculated radiological consequences resulting from these DBAs to ensure that 

the distances to the exclusion area boundary and the outer radius of the low 

population zone for a nuclear power plant, in conjunction with the operation 

of the dose-mitigating engineered safety features (ESF) systems, are 

sufficient to provide reasonable assurance that the dose reference values in 

10 CFR Part 100 will not be exceeded. 

For environmental qualification of electrical equipment, 10 CFR 50.49 states 

that safety-related electrical equipment, certain non-safety-related 

electrical equipment, and certain post-accident monitoring equipment should 

remain functional during and following design-basis events. Design-basis 

events are further defined as conditions of normal operation, including 

anticipated operational occurrences, DBAs, external events, and natural 

phenomena that the plant must be designed and built to withstand. 

In Regulatory Guide 1.89 (Revision 1), "Environmental Qualification of Certain 

Electrical Equipment Important to Safety for Nuclear Power Plants" (June 

1984), the staff assumes the TID-14844 source term for determining the 

radiation environment for qualification of electrical equipment important to 

safety and for calculating the time integrated radiation doses in the 

containment and at a point just above the containment sump. 

For additional Three Nile Island (TMI)-related requirements, 10 CFR 50.34 

refers to the TID-14844 source term (1) to perform radiation and shielding 

design review, (2) to assess post-accident sampling capability, (3) to assess 

the system design for leakage control and detection outside the containment, 

and (4) to evaluate control room habitability under accident conditions. 

Draft NUREG-1465 lists five fission-product release phases of a severe LWR 

accident: (1) coolant activity release, (2) gap activity release, (3) early 

in-vessel release, (4) ex-vessel release, and (5) late in-vessel release. The 

coolant activity release phase begins with a postulated pipe rupture and ends 

when the first fuel rod has been estimated to fail. During this phase, the 

fission products released to the containment atmosphere are those associated 

with radioactive material in the reactor primary coolant during normal plant 

operation. Draft NUREG-1465 does not specify fission-product releases to the 

containment during this phase, since the amounts of the fission products 

released are small compared with later release phases. The gap activity 

release phase begins when fuel cladding failure commences and ends when the 

fuel starts to melt. This phase involves the release of that radioactive 

material that has collected in the gap between the fuel pellet and cladding. 

The early in-vessel release phase begins when fuel and other materials melt 

and fall to the bottom of the reactor pressure vessel. During this phase, 

significant quantities of the volatile radionuclides in the core inventory, as 

well as some of the less volatile radionuclides, are estimated to be released 

into the containment. This release phase ends when the bottom head of the 

reactor pressure vessel fails, allowing molten core debris to fall onto the 

concrete below the reactor pressure vessel.
3

The ex-vessel release phase begins when the molten core debris exits the 

reactor pressure vessel and the core-concrete interaction takes place. This 

phase ends when the debris has cooled sufficiently so that significant quantities of fission products are no longer being released. During this 

phase, significant quantities of the volatile radionuclides not already released during the early in-vessel phase as well as lesser quantities of non

volatile radionuclides are released into the containment. Finally, the late 

in-vessel release phase commences at vessel breach and proceeds simultaneously with the ex-vessel release phase. 

Tables I and 2 of Attachment 2 list the proposed fission-product release 

fractions from proposed final NUREG-1465 of core inventory into the 

containment, including their release timing and duration. 

The staff proposes to use only the gap release and the early in-vessel release 

for design-basis calculations for the evolutionary and passive LWR designs. These source terms encompass a broad range of accident scenarios, including significant levels of core damage with the core remaining in the vessel. 

These would be the most severe scenarios from which the plant could be 

expected to return to a safe shutdown condition. Table 3 in Attachment 2 

shows that the NUREG-1465 release fractions associated with these releases are 

generally comparable to those of TID-14844 for the noble gases and iodine, but 

include additional nuclides, such as cesium. 

The staff considers the inclusion of the ex-vessel and late in-vessel source 

terms to be unduly conservative for DBA purposes. Such releases would only result from core damage accidents with vessel failure and core-concrete 

interactions. For evolutionary and passive LWRs, the estimated frequencies of 

such scenarios are low enough that they need not be considered credible for 

the purpose of meeting 10 CFR Part 100. 

The staff has addressed Issue I by selectively using the source terms given in 

proposed final NUREG-1465 using only "Gap Release" and "Early In-Vessel 

Release" (excluding "Ex-Vessel Release" and "Late In-Vessel Release" 

associated with vessel failure, and core-concrete interaction) in evaluating radiological consequences for DBAs, in evaluating the DBA radiation , environmental qualification of electrical and mechanical equipment important 

to safety, and in evaluating TMI-related requirements (for System 80+, AP600, 

and SBWR). 

ISSUE 2: Iodine Chemical Form (for System 80+, AP600, and SBWR) 

Regulatory Guides 1.3 and 1.4 specify that fission-product releases into the 

containment consist of 100 percent of the core inventory of noble gases and 50 

percent of iodines (half of which are assumed to deposit on interior surfaces 

of the containment very rapidly). The iodine chemical form is specified to be 

predominantly elemental iodine (91 percent), with 5 percent assumed to be 

particulate iodine and the remaining 4 percent assumed to be in the organic form. The 1 percent of "solid" fission products included in TID-14844 was 

dropped from consideration in Regulatory Guides 1.3 and 1.4.
4

In draft NUREG-1465, the staff concluded that iodine entering the containment 

from the reactor core is composed of at least 95 percent cesium iodide (CsI) 

with no more than 5 percent of iodine (1) and hydrogen iodide (HI). Once 

within the containment, highly soluble cesium iodide will readily dissolve in 

water pools forming iodide (Y) in solution and deposit onto the interior 

surfaces. The staff also stated in draft NUREG-1465 that the radiation

induced conversion of iodide (Y) in water into elemental iodine (12) is 

strongly dependent on the pH. The staff indicated that without pH control, 

large fractions of iodine dissolved in water pools in ionic form will be 

converted to elemental iodine and will be released into the containment 

atmosphere if the pH is less than 7. On the other hand, if the pH is 

maintained above 7, very little (less than 1 percent) of the dissolved iodine 

will be converted to elemental iodine. The EPRI requirements documents for 

evolutionary and passive plant designs and all evolutionary and passive LWR 

vendors require that the pH of the water in the containment be maintained 

above 7 (alkaline state) for the entire duration of the accident to minimize 

the formation of elemental iodine in the containment water in order to reduce 

the subsequent release of iodine into the containment atmosphere. The staff 

agrees with this requirement. 

In draft NUREG-1465, the staff did not address the formation of organic .iodide 

in the containment following an accident. However, organic iodide can be 

produced primarily by the reaction of elemental iodine with organic materials 

present in the containment. EPRI proposed the use of 0.15 percent organic 

iodide based on 3 percent conversion of elemental iodine to organic iodine 

(i.e., 3 percent of 5 percent is 0.15 percent). Based on studies by NRC 

contractors documented in NUREG/CR-4461, -5232, and -5950, the staff estimates 

that no more than 3 percent of the airborne elemental iodine will be converted 

into organic species. This amount of organic iodide would thus correspond to 

about 0.15 percent of the core iodine inventory (i.e., 3 percent conversion of 

5 percent elemental iodine is 0.15 percent). Final NUREG-1465 will address 

this issue. 

The final NUREG-1465 will not alter the staff's conclusion in the System 80+ 

FSER in regard to its acceptance of the System 80+ design for meeting the dose 

reference values given in 10 CFR Part 100 and the control room operator dose 

limit specified in GDC 19. As shown in Table 3 of Attachment 2, the fission

product release fractions given in final NUREG-1465 are the same (for noble 

gases, iodine, and cesium) or lower than those given in draft NUREG-1465 (for 

non-volatile elements). For fission-product chemical forms, the percent of 

core iodine inventory in organic form is reduced to 0.15 in final NUREG-1465 

from 0.25 percent used in CE System 80+ design review. There were no changes 

made in fission-product release timing for PWR in final NUREG-1465 from that 

in draft NUREG-1465. 

The staff has addressed Issue 2 by partitioning radioiodine chemical forms as 

follows: 95 percent cesium iodide (CsI, or I- in the aqueous phase); 4.85
5

percent elemental iodine'(12), and 0.15 percent as organic iodide (e.g., CH3I). The latter two values will be used In preference to a single value of 

5 percent 12 recommended In draft NUREG-1465. These values for the iodine 

chemical form will be used in the AP600 and SBWR reviews. 

The evolutionary and passive LWR vendors have stated that they will use the 

staff's proposed values until final NUREG-1465 is published. The staff will 

use these chemical forms in evaluating the containment spray removal 

efficiencies, BWR suppression pool decontamination factors, passive deposition 

of fission products in the containment, control room habitability, and 

engineered safety features (ESF) filtration systems in its evolutionary and 

passive LWR design reviews. The application of these chemical forms in each 

evolutionary and passive LWR design review will be discussed in the staff's 

forthcoming safety evaluation reports. 

ISSUE 3: Equipment Survivability for Design Features Needed for Severe 

Accident Mitigation and Containment Integrity (for System 80+, AP600, and 

SBWR). 

As the staff concluded in SECY-90-016, the equipment and features provided for 

preventing and mitigating severe reactor accidents need not be subject to the 

rigorous 10 CFR 50.49 environmental qualification requirements. However, 

mitigation features and post-accident sampling systems for severe reactor 

accidents must be designed to provide a reasonable level of confidence that 

they will survive the severe accident environment for which they are intended 

and be able to operate over the timespan for which they are needed. 

For the purpose of radiological assessments, the staff proposes to define the 

DBA radiation environment as that resulting from fission-product releases from 

coolant activity release, gap release, and in-vessel release. The staff is 

also defining the severe reactor accident radiation environment as that 

resulting from the above fission-product releases plus ex-vessel release and 

late-in-vessel release that are associated with vessel failure and core

concrete interaction. 

If safety-related equipment is relied on to cope with severe accident 

situations, there should also be a reasonable level of confidence that this 

equipment will survive the severe accident environment, including radiation, 

for the period that it is needed to perform its intended function. The staff 

will evaluate the evolutionary and passive LWR vendors' identification of such 

equipment and features, their intended functions, location, arrangement, 

shielding, and the period during which they are needed to perform on a case

by-case basis. 

For issues related to severe accidents, such as the survivability of equipment 

needed for severe accident mitigation and post-accident sampling, the staff 

believes that the ex-vessel and late-in-vessel releases can represent an 

important part of the source terms. However, for those evolutionary and 

passive designs that include reliable ex-vessel flooding mechanisms for the 

core debris, the ex-vessel contribution may be significantly reduced by 

aerosol scrubbing by a water pool overlying core debris.
6

The staff has addressed this part of Issue 3 by requiring that equipment and 

features needed for severe accident prevention, mitigation, and post-accident 

sampling be designed to provide a reasonable level of confidence that they 

will operate In a severe accident environment. This environment would Include 

the ex-vessel release, with proper credit for features to mitigate that 

release, and the late In-vessel release (for CE System 80+, AP600, and SBWR). 

ISSUE 4: Radioactive Iodine Deposition on BWR Main Steamlines and Condensers 

(for ABWR and SBWR). 

The main steamlines in BWR plants, including the ABWR and SBWR, contain dual 

quick-closing main steam isolation valves (MSIVs). These valves function to 

isolate the reactor system in the event of a break in a steamline outside the 

primary containment, a design-basis LOCA, or other events requiring 

containment isolation. Although the MSIVs are designed to provide a leaktight 

barrier, it is recognized that some of the valves will leak. Operating 

experience has indicated that the leaktightness of MSIVs has occasionally 

degraded and the leakage limits in the technical specifications have not 

always been maintained. 

When calculating offsite consequences of potential accidents, the staff 

conservatively assumed that radioactive material passing through the MSIV at 

its technical specifications leakage limit is released directly into the 

environment. No credit was given for the integrity and leaktightness of the 

main steam piping and condenser to provide holdup and plateout of fission 

products. 

Because of recurring problems with excessive leakage of MSIVs, the staff 

recommended installation of a supplemental leakage control system (LCS) in 

Regulatory Guide 1.96, "Design of Main Steam Isolation Valve Leakage Control 

Systems for Boiling Water Reactor Nuclear Power Plants,' to ensure that the 

isolation function of the MSIVs is in accord with the specified limits. Most 

of the operating BWRs have an LCS. 

In response to the MSIV leakage concerns, the BWR Owners Group commissioned a 

program of studies to determine the causes of high leak rates and the means to 

eliminate them. The results of these studies were submitted to the staff in 

General Electric (GE) proprietary reports, NEDO-31643P (November 1988) and 

NEDO-31858P (February 1991), both entitled "Increasing Main Steam Isolation 

Valve Leakage Rate Limits and Elimination of Leakage Control Systems., 

EPRI and GE referenced these reports in the requirements documents and the 

ABWR and SBWR standard safety analysis reports (SSARs), respectively, as the 

basis for not providing arLLjS and for requesting a substantially higher MSIV 

total leak rate [1.1 x 10" m /sec (140 ft/hr) rather than 3.5 x 10" m/sec 

(45 ft /hr)] than the typically allowed limit for operating BWRs. In 

addition, GE claimed in the ABWR and SBWR SSARs, that elemental and 

particulate iodine will deposit on the main steamlines (MSLs) and condenser, 

thus mitigating the radiological consequences of an accident. The basis for 

GE's claim was contained in a model developed for this purpose and documented 

in the referenced NEDO reports.
7

Following a LOCA, three potential release pathways exist for main steam 

leakage through the MSIVs: 

(1) main steam drainlines to the condenser with delayed release to the 

environment through the low-pressure turbine seals 

(2) turbine bypass lines to the condenser with delayed release to the 

environment through the low-pressure turbine seals 

(3) main steamline turbine stop and control valves through the high-pressure 

turbine seals to the environment bypassing the condenser 

The consequences of leakage from pathways 1 and 2 will be essentially the same 

because the condenser can be used to process MSIV leakage. The iodine removal 

efficiency of the condenser will vary depending on the inlet location of the 

bypass or drainline piping, but in either case, iodine will be removed. For 

pathway 3, MSIV leakage through the closed turbine stop and control valves 

will not be processed via the condenser. For this case, the high-pressure 

turbine (having a large internal surface area associated with the turbine 

blades) will remove iodine. 

The staff believes that as long as either the turbine bypass or drainline 

leakage pathway is available, MSIV leakage through the closed turbine stop and 

control valves (pathway 3) will be negligible. Essentially all of the 

releases will be through the main condenser because there will be no 

differential pressure in the HSL downstream of the MSIVs following the closure 

of the valves. 

Furthermore, MSIV leakage through pathway 3, if any, will have been subjected 

to the same iodine-removal processes in the MSLs (up to turbine stop valves) 

as those in the other pathways. The leakage will be further subjected to 

iodine removal by deposition in the high-pressure turbine internal surfaces. 

Removal by the main condenser is not applicable in pathway 3. 

Basic principles of chemistry and physics predict that gaseous iodine and 

airborne iodine particulate material will deposit on surfaces. Several 

laboratory and in-plant studies have demonstrated that gaseous iodine will be 

deposited by chemical adsorption, and particulate iodine will be deposited 

through a combination of sedimentation, molecular diffusion, turbulent 

diffusion, and impaction. The transport of gaseous iodine in elemental and 

particulate forms also has been studied for many years, and several groups 

have proposed different models to describe the observed phenomena. 

The staff has developed a model that treats the MSIV leakage pathway as a 

sequence of small segments for which instantaneous and homogeneous mixing is 

conservatively assumed; the mixing computed for each segment is passed along 

as input to the next segment. The number of segments depends on the 

parameters of the line and flow rate and can be as many as 100,000 for a long, 

large-diameter pipe and a low flow rate. Each line segment is divided into 

five compartments that represent the concentrations of the three airborne 

iodine species, the surface that contains iodine available for resuspension, 

and surface iodine that has reacted and is fixed on the surface.
8 

The staff's model considers three iodine species: elemental, particulate, and 

organic. A fourth species, hypoiodous acid, was considered for the purpose of 

the staff's model to be a form of elemental iodine. All iodine in the segment 

undergoes radioactive decay. The resulting concentration from each segment of 

the deposition compartment serves as the input to the next segment. The 

technical references indicate that particulate and elemental iodine would be 

expected to deposit on surfaces with rates of deposition varying with 

temperature, pressure, gas composition, surface material, and particulate 

size. Therefore, the staff believes that an appropriate credit for the 

removal of iodine in the MSLs and main condensers should be provided in the 

radiological consequence assessment following a design-basis accident. 

In the review of the ABWR design, the staff accepted GE's proposed elimination 

of the LCS and allowed a higher MSIV leakage rate providing an appropriate 

credit for the removal of iodine in the MSLs and condenser. The staff gave a 

detailed technical evaluation in the ABWR draft final safety evaluation report 

(DFSER). Even though the staff used the TID-14844 source terms (mostly 

elemental iodine) for the ABWR design, as proposed by GE, the staff's model is 

also applicable for the NUREG-1465 source terms (mostly iodine in particulate 

forms as aerosol). 

The GE model, as well as the one used by the staff, is based on time-dependent 

temperature adsorption phenomena with instantaneous and perfect mixing in a 

given volume. Both models use the same MSIV leakage pathways. They differ, 

however, in the treatment of buildup of iodine in the tSLs and condenser. GE 

assumed steady-state iodine in equilibrium in a large volume, while the staff 

assumed transient buildup of iodine in a finite number of small volumes. The 

staff does not consider these differences to be significant and finds that the 

resulting radiological consequences (offsite doses) are in good agreement. It 

therefore finds the GE model to be acceptable. 

Sections III(c) and VI of Appendix A to 10 CFR Part 100 require that 

structures, systems, and components necessary to ensure the capability to 

mitigate the radiological consequences of accidents that could result in 

exposures comparable to the dose guidelines of Part 100 be designed to remain 

functional during and after a safe shutdown earthquake (SSE). Thus, the MSL, 

portions of its associated piping, and the main condenser are required to 

remain functional if credit is taken for deposition of iodine and if the SSE 

occurs. Consequently, the staff's past practice has been to classify these 

components as safety related and seismic Category I. In addition, Appendix A 

to 10 CFR Part 100 requires that the engineering method used to ensure that 

the safety functions are maintained during and after an SSE involves the use 

of either a suitable dynamic analysis or a suitable qualification test. 

In Section II.E, 'Classification of Main Steamlines in Boiling Water 

Reactors," of SECY-93-087, the staff stated that the classification of the MSL 

in the turbine building as non-seismic Category I is needed for consistency 

with the classification of the turbine building. On this basis, the quality 

and safety requirements imposed on the MSL from the outmost isolation valve up 

to, but not including, the turbine stop valve are equivalent to the staff 

guidelines in Appendix A to SRP Section 3.2.2, "System Quality Group 

Classification," and Regulatory Guide 1.29, "Seismic Design Classification."
9

In SECY-93-087 (Item II.E), the staff proposed for Commission approval, among other things, that: 

(1) neither the main steam drain and bypass lines from the first valve up to 

the condenser inlet nor the piping between the turbine stop valve and 

the turbine inlet be classified as safety related or as seismic Category 1; rather, these lines should be analyzed using a dynamic seismic 

analysis to demonstrate structural integrity under SSE loading 

conditions; 

(2) the turbine stop, control, and bypass valves and the NSLs from the 

turbine control valves to the turbine meet all of the quality group and 

quality assurance guidelines specified in Appendix A to SRP Section 

3.2.2; and 

(3) seismic analyses be performed to ensure that the condenser anchorages and the piping inlet nozzle to the condenser are capable of maintaining their structural integrity during and after the SSE. 

The staff concluded in SECY-93-087 that the above-described approach provides reasonable assurance that the main steam piping from the outmost isolation 

valve up to the turbine stop valve, the main steam drain and bypass lines up 

to the condenser, and the main condenser will retain their pressure and 

structural integrity during and following SSE. The Commission approved the 

above staff proposals in its staff requirements memorandum of July 21, 1993. 

In its review of the ABWR design, the staff concluded in the FSER that the 

ABWR design provides reasonable assurance that the main steam piping from the 

outmost isolation valve up to the turbine stop valve, the MSIV leakage pathway (i.e., the drainline or bypass line) up to the condenser, and the main 

condenser will remain structurally intact and leaktight, so that they can act 

as a holdup volume for fission products during and following an SSE. 

The staff also determined that the ABWR design meets the requirements of 

10 CFR Part 100 because the structures, systems, and components described 

above are designed to remain functional during and following an SSE. 

For the purpose of providing a credit for iodine holdup and plateout, the 

staff's model requires that the main steam piping (including its associated 

piping to the condenser) and the condenser remain structurally intact 

following an SSE, so they can act as a holdup volume for fission products. By the term "structurally intact," the staff assumes the steamline will retain 

sufficient structural integrity to transport the MSIV bypass leakage at its 

relatively low flow rate throughout the steamlines and condenser. The staff 

considers, in its radiological consequence assessment, that the condenser is 

open to the atmosphere via leakage through the low-pressure turbine seals. 

Thus, it is only necessary to ensure that gross structural failure of the 

condenser will not occur. 

The staff has addressed Issue 4 by accepting the passive BWR plant designs without an LCS and allowing an appropriate credit for iodine removal in the 

HSL and condenser following a design-basis accident (for ABWR and SBWR).
10

The robustness of the leak pathways and the amounts of iodine removed are 

plant design specific and, therefore, will be determined on a case-by-case 

basis. 

ISSUEL: Fission-Product Holdup in the Secondary Containment (for SBWR). 

The structure or structures that completely surround the reactor primary 

containment and that are held at a negative pressure can be classified as a 

secondary building for the purpose of fission-product control and, therefore, 

for the mitigation of radiological consequences. The EPRI requirements 

documents for evolutionary and passive plans designs require no holdup and 

retention of fission products in the PWR secondary building. Westinghouse has 

not claimed any credit for fission-product holdup in the AP600 auxiliary 

(secondary) building. 

GE proposed a safety envelope design for the SBWR plant. This envelope is a 

reinforced concrete structure (secondary containment) within the reactor 

building that forms an envelope completely surrounding the primary containment 

(except the basement drywell head and drywell top slab). During normal 

operation, the safety envelope is maintained at a slightly negative pressure 

relative to the atmosphere by the reactor building heating, ventilation, and 

air conditioning system. The safety envelope is designed to be automatically 

isolated on detection of, among other things, a high radiation environment. 

Its design leakage rate is not to exceed 25 percent of the safety envelope 

free volume per day at a differential pressure of 6 mm (0.25 inch) of water. 

GE stated in the SBWR standard safety analysis report that the safety envelope 

is designed to be capable of periodic testing to ensure its intended 

performance and that testing and inspection of the integrity of the safety 

envelope, including its leakage rate, will be performed in accordance with the 

specific plant technical specifications. To provide a credit for fission

product holdup, the staff will require periodic leak rate testing similar to 

that performed for the primary containment. For safety envelope air mixing, 

GE proposed 50 percent on the basis of one-dimensional modeling of the KEMA 

computer code using simplified compartment volumes. GE informed the staff 

that it will follow this simplified modeling with a three-dimensional modeling 

effort using a computational fluid dynamic computer code to better define the 

capability of the SBWR safety envelope to hold up radioactive material and 

mitigate any potential leakage. GE further informed the staff that this 

effort will be completed in 1994. 

The staff has addressed Issue 5 by allowing appropriate fission-product holdup 

(for decay) credit without requiring that a negative pressure be maintained in 

the SBWR secondary containment (building) If (1) the vendor specifies that the 

secondary containment leakage and mixing performance be consistent with the 

values used by the staff in Its radiological assessment and (2) the COL 

combined license applicant Incorporates the secondary containment leakage 

value specified by the vendor into the plant-specific technical specifications 

(for SBWR).
11

ISSUE 6: Fission-Product Release Timing (for System 80+, AP600, and SBlR) 

In SECY-90-309, "Impacts of Source Term Timing on NRC Regulatory Positions," the staff informed the Commission of its plans to change regulatory positions for both existing and future advanced reactor designs that result from more 

realistic treatment of fission-product release timing. In Regulatory Guides 

1.3 and 1.4, the instantaneous release of fission products into the 

containment on receipt of a DBA signal from maximum full-power operation of 

the core is postulated and fission products released are assumed to be 

immediately available for leakage from the containment atmosphere to the 

environment. 

The staff evaluated a number of items in SECY-90-309 to determine if more 

realistic treatment of fission-product releases would lead to less demanding regulatory requirements. The staff concluded that relaxation of requirements 

was justified for (1) the containment purge/vent isolation valve closure time, 

(2) the main steam isolation valve closure time, and (3) the emergency diesel 

generator start time. It also concluded that relaxing the timing requirements for each of these items would affect more than radiological consequences and, 

therefore, a plant-specific safety analysis would be required for each 

proposed relaxation. 

In draft NUREG-1465, the staff indicated that fission-product gap activity release was estimated for a large-break LOCA to commence no earlier than 10 to 

30 seconds for PWRs and 30 seconds for BWRs. It also indicated that fission

product early in-vessel release was estimated to commence no earlier than 0.5 

and 1.0 hour for PWRs and BWRs, respectively. In contrast to the 

instantaneous releases that were postulated in Regulatory Guides 1.3 and 1.4, analyses of severe accident sequences have shown that fission-product releases 

can be generally categorized in terms of phenomenological phases associated 

with fuel failure, melting, and relocation despite differences in reactor 

design and accident sequence. 

The single values for release timing and durations shown in draft NUREG-1465 

have been chosen on the basis of simplicity, consistency, and conservatism. 

An accurate determination of the release timing and durations will depend not 

only on the reactor design but also on the selected accident sequences. The 

staff will review an evolutionary or passive reactor vendor's design and , accident sequence as a basis for fission-product release timing to determine 

if its'values are different from those provided in draft NUREG-1465. For its 

reviews of evolutionary and passive LWR designs, the staff will use the 

fission-product release timing values for the radiological consequence 

assessments of (1) DBAs, (2) the operation of control room habitability 

systems, (3) the operation of ESF filtration systems, (4) containment and MSIV 

closure, (5) containment purge/vent isolation valve closure, and (6) emergency diesel generator start time. 

In draft NUREG-1465, the staff stated that the coolant activity release starts 

at the beginning of the accident, the gap activity release at no earlier than 

10 minutes into the accident (assuming approval of leak-before-break 

methodology by the staff), and the early in-vessel release no earlier than 30 

and 60 minutes into the accident for PWRs and BWRs, respectively. This timing
12 

is based on a large-break LOCA of current LWR design. The staff realizes that 

fission-product release timing is dependent upon reactor type, design, and the 

bounding reactor design basis accident sequences chosen for the source term 

application. 

Therefore, the staff will evaluate each proposed ALWR design on a case

specific basis to determine appropriate fission-product release timing. The 

staff will discuss the application of fission-product release timing in each 

evolutionary and passive LWR design review in its forthcoming safety 

evaluation reports. 

ISSUEZ: Aerosol Deposition in Primary Containment (for AP600 and SBWR) 

The principal means of removing the airborne fission products from the 

containment atmosphere in the LWR design traditionally included use of active 

containment atmosphere cleanup systems such as containment spray, ESF 

filtration, and pressure-suppression pool scrubbing. In the TID-14844 source 

terms, the staff assumed that 50 percent of radioactive iodine (91 percent in 

the form of elemental iodine) released into the containment was immediately 

removed by plateout (diffusion mechanism). No fission products in aerosol 

form were assumed to be present. In the passive LWRs, an active containment 

atmosphere cleanup system is not provided. Reliance is placed solely on such 

natural aerosol removal processes in the containment as holdup (for decay), 

sedimentation (for settling), diffusion (for plateout), and leakage (for 

depletion). 

In response to the staff's request for additional information, EPRI stated 

that because a containment spray system is not used for the passive plant, the 

means by which aerosol Is removed will be dominated by such natural removal 

effects as diffusion and sedimentation. 

The most complete mechanistic treatment of aerosol behavior in the reactor 

containment is found in CONTAIN, a computer code developed at Sandia National 

Laboratories under NRC sponsorship for the analysis of containment aerosol 

behavior under severe accident conditions. The other computer code, NAUA, 

which is very similar to the aerosol portion of CONTAIN, was developed at 

Kernforschungszentrum, Karlsruhe, Federal Republic of Germany, and was used 

for aerosol treatment In the NRC Source Term Code Package. There are a number 

of other aerosol behavior computer codes, but these two codes are the most 

widely used and accepted throughout the international nuclear safety 

community. Either code is acceptable to the staff for the evaluation of 

aerosol fission-product behavior in the primary containment following a severe 

reactor accident. However, most of these computer codes use lumped parameter 

methods, which assume that the aerosols are well mixed in the containment. 

When the condition of mixing cannot be established, the calculated results 

using the above computer codes may not be valid. 

The staff will consider two natural processes for removing aerosols from the 

containment atmosphere over the entire period of an accident (30 days): (1) 

sedimentation mechanism of gravitational settling, including aerosol 

agglomeration, and (2) diffusion mechanisms of diffusiophoresis and 

thermophoresis. Diffusiophoresis is the diffusion of particles into interior
13

heat sink surfaces, such as the containment walls on which steam is condensing. Thermophoresis is a similar effect due to a temperature gradient 

near the surface. 

In consideration of these two natural processes for removing aerosols from the containment atmosphere, the staff will evaluate the thermal hydraulic conditions and production of nonradioactive aerosols since they can strongly influence the behavior of radioactive aerosols in containment following a DBA. The thermal hydraulic conditions include the containment pressure, relative humidity, and steam condensation and heat removal rates of the containment structure. These thermal hydraulic parameters, as well as the amounts of non radioactive aerosol produced, differ with the specific reactor accident 

sequences and with the accident mitigation features provided (e.g., isolation condensers and primary containment cooling system for the SBWR design, containment shell cooling for the AP600 design, and containment spray for the CE System 80+ design). 

In a paper on the source term for passive plants supplemented by a draft report titled "Passive ALWR Containment Natural Aerosol Removal" (April 1993), EPRI has provided radioactive aerosol removal rates based on sedimentation and diffusion of aerosols in the containment atmosphere using the EPRI version of the NAUA code (NAUAHYGROS). 

The staff agrees with EPRI on the physical processes associated with natural aerosol removal that could be taken into account in establishing the airborne fission-product concentrations in the containment atmosphere. Therefore, the staff accepts the aerosol removal mechanisms proposed by EPRI in the requirements documents for evolutionary and passive plant designs. However, the containment aerosol removal rates are plant design specific and will vary, depending on, but not limited to, the containment geometry, containment size, surface area, steam quality, and containment cooling mechanisms. 

Because the AP600 and SBWR designs have been submitted and are under review, the staff does not intend to develop and promulgate specific guidance relative to the impact of thermal hydraulic conditions and production of nonradioactive aerosols on the behavior of radioactive aerosols in containment at this time. The staff will, however, communicate on these matters with Westinghouse and General Electric via Requests for Additional Information (RAI) or similar licensing communication vehicles during the development of the draft safety evaluation reports. In the discussion which follows, the staff has summarized the current status of the three ALWR design reviews which are based on the 

revised accident source terms. 

General Electric proposed a group of reference accident sequences for the SBWR design that lead to core damage but terminate with the reactor pressure vessel still intact to determine the thermal hydraulic conditions in the containment following the design basis accident. GE, in turn, used these reference accident sequences to generate the thermal hydraulic conditions using an SBWR version of the KAAP code. The staff is currently reviewing the GE proposal and will either select its own representative group of accident sequences or, after appropriate review, use the GE-selected accident sequence for the staff's independent determination of the thermal hydraulic conditions and
14

nonradioactive aerosol generation. The staff will use this information along 

with the following parameters for evaluating the fission-product behavior and 

determining the aerosol removal rates in the SBWR containment following a 

design-basis accident: 

1) containment geometry 

2) aerosol characteristics 

3) aerosol removal by isolation condensers and primary 

containment cooling systems 

Westinghouse has neither provided design basis accident sequences nor the 

thermal hydraulic conditions in the containment for the AP600 design at this 

time. The staff is currently addressing this matter with Westinghouse via 

RAIs. 

The CE System 80+ evolutionary plant design includes a safety-grade 

containment spray system; therefore, the thermal hydraulic conditions and the 

amounts of nonradioactive aerosol are less significant for the purpose for 

determining radioactive aerosol behavior-and its removal rates in the 

containment. Nevertheless, the staff used in its evaluation, a heat removal 

rate of 100 moles per second (typical value from NUREG-1150) and 350 kilograms 

of nonradioactive aerosol (a value given in draft NUREG-1465 as information) 

in its evaluation of the System 80+ design. 

ISSUE 8: Aerosol Removal by BWR Suppression Pool (for ABWR and SBWR) 

The BWR suppression pools are designed primarily as containment pressure and 

temperature suppression mechanisms for reactor pressure vessel blowdown. 

However, they can also serve as a medium for scrubbing radioactive fission 

products except noble gases and iodine in organic forms. The scrubbing 

(attenuation) of fission products in suppression pools is usually expressed as 

a "decontamination factor" (DF), which is defined as the ratio of the 

radioactive material injected into the pool divided by the airborne 

radioactive material that leaves the surface of the pool water. 

Regulatory Guide 1.3, issued in 1974, does not recommend credit for fission

product scrubbing by BWR suppression pools. However, the Reactor Safety Study 

(WASH-1400), issued in 1975, assumed a DF of 100 for subcooled suppression 

pools and 1.0 for steam-saturated pools. Since 1975, the staff and EPRI have 

developed detailed models for the analysis of radioactive aerosol scrubbing by 

the suppression pool. Accordingly, in 1988, the staff issued revised SRP 

Section 6.5.5, "Pressure Suppression Pool as a Fission Product Cleanup 

System," stating that ignoring the large volume of research data supporting 

pool credit would be an undue degree of conservatism. The revised SRP section 

allowed a maximum DF of 10 for particulate and elemental iodines for Mark II 

and Mark III containments and a DF of 5 for a Mark I containment. These DF 

values were based on the TID-14844 report, which stipulates an instantaneous 

release of fission products from the reactor coolant system to the containment 

atmosphere. In the revised SRP section, this was further subjected-to the 

fraction of the drywell atmosphere bypassing the suppression pool by leaking 

through drywell penetrations to obtain the overall DF.
15

GE utilized the TID-14844 source terms for its ABWR design. The staff 

conservatively assumed in its radiological consequence assessment a DF of 2 by 

the ABWR suppression pool (equivalent to suppression pooi steam bypass of 50 

percent) for airborne radioactive iodine in elemental and particulate forms. 

The staff has accepted the drywell leakage values [0.005 m (0.05 ft 2 ) of 

effective leakage area pathway] proposed by GE for the ABWR. The ABWR 

containment sprays in the drywell or wetwell, or both, also would reduce the 

effect of suppression pool bypass leakage on containment performance. 

Therefore, a DF of 2 used in the staff's analysis is conservative. For the 

ABWR design, the staff will review the plant-specific technical 

specifications, which require periodic inspections to confirm suppression pool 

depth, and surveillance tests to confirm drywell leaktightness on a case-by

case basis. 

ISSUE9: Fission-Product Removal by Containment Spray (For AP600) 

General Design Criteria (GDC) 41, 42, and 43 of Appendix A to 10 CFR Part 50 

require systems to control fission products to reduce the concentration that 

may be released to the environment. The containment spray system reduces 

containment pressure and temperature and removes airborne radioactive fission 

products in the containment atmosphere following a LOCA. 

The EPRI requirements document for evolutionary plant designs requires a 

containment spray system. ABB-CE has provided an acceptable spray system for 

the System 80+ design. For its ABWR design, GE provided a safety-related 

spray system in the containment to reduce pressure and temperature following a 

LOCA, but has not claimed nor has the staff provided, any credit for the 

removal of airborne fission products for the purpose of radiological 

consequence assessments. 

The EPRI requirements document for passive plant designs requires neither an 

active containment atmosphere cleanup system nor a containment spray system. 

This document states that fission-product control should be accomplished 

passively by reliance on natural deposition (plateout) and holdup (decay) and 

slower fission-product release timing. The Westinghouse AP600 design relies 

on plateout and holdup in the primary containment. In the SBWR design, GE 

relies on plateout in the reactor primary containment and holdup in the 

reactor primary containment and in the safety envelope. The safety envelope 

is a reinforced-concrete structure within the reactor building that forms an 

envelope completely surrounding the containment. It is designed to be 

periodically tested for leaktightness. The SBWR design includes a non-safety

grade spray system for both the upper drywell and suppression chamber. 

However, GE has not claimed, nor has the staff provided, fission-product 

removal credit for the non-safety-grade spray system in the SBWR design. The 

AP600 design does not provide a non-safety-grade containment spray. 

The passive LWR vendors have submitted radiological consequence assessments to 

demonstrate that without containment spray, the passive LWR designs can meet 

the dose reference values of 10 CFR Part 100 and the control room operator 

dose limits in GDC 19 using the fission-product release timing based on the 

specific designs and selected accident sequences and using their best estimate 

of fission-product removal efficiencies. Unlike the current generation of
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operating PWRs, the Westinghouse AP600 design does not include containment 

sprays. The staff will review the AP600 radiological consequence assessment 

to determine if the dose reference values in 10 CFR Part 100 and the control 

room operator dose criteria of GDC 19 can be met without a containment spray 

system. That determination will involve evaluation of aerosol removal rates 

by natural deposition, containment leak rates, and assurance of pH control. 

ISSUE 10: Radioactive Aerosol and Iodine Removal by Engineered Safety 

Features (ESF) Atmosphere Cleanup System (for AP600 and SBWR) 

Containment atmosphere cleanup systems are to be provided as necessary to 

reduce the amount of radioactive material released to the environment 

following a postulated design-basis accident in accordance with GDC 41, 42, 

and 43. These GDC also require that these systems be designed to permit 

appropriate periodic inspection and testing to ensure their integrity, 

capability, and operability. 

The EPRI requirements document for passive plant designs does not require an 

ESF atmosphere cleanup system, and the passive LWR vendors have not provided 

these systems in the AP600 and SBWR designs. The System 80+ design Includes 

an ESF atmosphere cleanup system with charcoal adsorbers for the annulus 

building ventilation system. However, ABB-CE did not claim any fission

product removal credit for charcoal adsorbers; therefore, they will not be 

required to be in the CE System 80+ technical specifications. The GE ABWR 

design includes an ESF atmosphere cleanup system with HEPA filters and 

charcoal adsorbers for the reactor building (standby gas treatment system) and 

control room habitability system. Since the chemical forms of iodine will be 

predominantly in particulate form, HEPA filters rather than charcoal adsorbers 

will play a major role in removing airborne fission products following an 

accident. 

The passive LWR vendors have submitted the radiological consequence 

assessments to demonstrate that without an ESF atmosphere cleanup system, the 

passive LWR designs can meet the dose reference values of 10 CFR Part 100 and 

the control room operator dose limits in GDC 19 using the fission-product 

release timing based on its specific passive LWR design and selected accident 

sequences and using their best-estimate fission-product removal efficiencies. 

The staff will review these assessments and provide its conclusions in the 

safety evaluation reports. 

Therefore, the staff will not require an ESF atmosphere cleanup system, 

provided the passive LWR vendor demonstrates that the passive LWR design, 

including adequate pH control, can meet the dose reference values in 10 CFR 

Part 100 and the control room operator dose limits in GDC 19. 

ISSUE 11: Atmospheric Diffusion Model for Control Room Habitability 

Assessment (for CE System 80+, AP600 and SBWR) 

The staff has developed with contractor assistance a new model for determining 

atmospheric relative concentrations (XLQ) in building wakes to be used for 

control room habitability assessment. The new model is applicable for either 

TID-14844 or revised accident source terms and may be used by the evolutionary
17

and passive LWR vendors and existing licensees. The staff will use this model in its review of evolutionary and passive LWR designs using the given building layouts and standard plant site design parameters. : 

The current guidelines for determining x/Q are given in Regulatory Guides 1.3, 1.4, and 1.145, wAtmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," and in SRP Section 6.4, "Control Room Habitability Systems." The new model, in general, is more 

realistic in determining the atmospheric dispersion factors for assessing concentrations of radioactive material at the control room emergency air intake except for the reactor sites with a high average wind speed (e.g., sites around the Great Lakes in the United States). This model is not proposed for backfitting on existing plants. However, because it reflects the 

current understanding of building wake behavior, the staff plans to make this 

new model available to existing licensees who may voluntarily propose to use it in future license amendments. The staff will use this new model for 

reviewing future reactor designs and their siting. 

The new model differs from the current model in that it assumes the wind will 

transport radioactive materials directly from the release point to the control 

room air intake only if the wind direction is within ± 30 degrees of the direction needed to transport radioactive material directly to the intake. It also determines 2-hour and 4-hour average concentrations from hourly centerline X/Q values. It will use hourly, sector-average X/Q values to calculate average concentrations for periods of 24, 48, 96, 168, 360, and 720 hours. All averages will be computed as moving (overlapping) averages of the 

appropriate hourly X/Q values. 

In May 1994, NRC contractor (Pacific Northwest Laboratories) conducted a peer review group meeting for NRC to discuss and resolve, by consensus, any comments on this new model. The peer review group consisted of experts having demonstrated expertise in meteorology and atmospheric dispersion modeling. The peer group did not raise any major concerns, and the staff is considering their comments for incorporation into the model. 

The staff will use the new model in determining the atmospheric dispersion factors for assessing concentrations of radioactive material at the control 

room emergency air intake following a design-basis accident for evolutionary and passive LWR design reviews. 

ISSUE 12: Failure of Heat Exchanger Tubes in the Passive Containment Cooling System (for SBWR) 

The SBWR passive containment cooling system (PCCS) will remove the core decay heat rejected to the containment after a LOCA. It consists of three independent closed loops that are extensions of the containment. Each loop contains a heat exchanger that condenses steam on the tube side and transfers heat to the water in a large pool located outside the containment. The PCCS loops are operated at a low pressure and are initially driven by the pressure difference created between the containment drywell and suppression pool and then by gravity drainage of steam condensed in the tubes. Three PCCS loops 

are an extension of the containment and do not have isolation valves. The
18

potential for containment bypass in this manner only applies to the SBWR 

design. 

Should the PCCS heat exchanger tubes fail, the PCCS will provide a bypass 

pathway for the SBWR containment, releasing radioactive fission products from 

the containment atmosphere to the reactor building through the passive 

containment cooling (PCC) pool water. The staff expects the radioactive 

aerosol scrubbing by the PCC pool water to be similar to that by theSBWR 

pressure-suppression pools. 

The staff has not completed its review of the SBWR PCCS design, nor has it 

determined the important characteristics (such as carrier fluid conditions, 

failure configuration, pathway evaluation, and pool thermal and hydraulic 

characteristics) affecting the radioactive aerosol removal (source term 

behavior) by the PCC pool water. The staff's evaluation of this potential 

containment bypass with its source term behavior will be given in the safety 

evaluation report for the SBWR. The discussion in this section provides 

information relative to staff commitments identified in SECY-93-087.