ML21067A645

DPO-2020-002, Differing Professional Opinion (DPO) Case-File (Redacted Public Version)

ML21067A645
Person / Time
Issue date:	03/08/2021
From:	Ian Gifford NRC/OE
To:
Gifford I
References
DPO-2020-002
Download: ML21067A645 (69)

Text

DPO Case File for DPO-2020-002 The following pdf represents a collection of documents associated with the submittal and disposition of a differing professional opinion (DPO) from an NRC employee involving RG 1.183 Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors.

Management Directive (MD) 10.159, The NRC Differing Professional Opinions Program, describes the DPO Program. https://www.nrc.gov/docs/ML1513/ML15132A664.pdf The DPO Program is a formal process that allows employees and NRC contractors to have their differing views on established, mission-related issues considered by the highest level managers in their organizations, i.e., Office Directors and Regional Administrators. The process also provides managers with an independent, three-person review of the issue (one person chosen by the employee). After a decision is issued to an employee, they may appeal the decision to the Executive Director for Operations (or the Commission, for those offices that report to the Commission).

Because the disposition of a DPO represents a multi-step process, readers should view the records as a collection. In other words, reading a document in isolation will not provide the correct context for how this issue was reviewed and considered by the NRC.

It is important to note that the DPO submittal includes the personal opinions, views, and concerns by NRC employees. The NRCs evaluation of the concerns and the NRCs final position are included in the DPO Decision.

The records in this collection have been reviewed and approved for public dissemination.

Document 1: DPO Submittal Document 2: Memo Establishing DPO Panel Document 3: DPO Panel Report Document 4: DPO Decision

Document 1: DPO Submittal NRC FORM680 r;J,""E:'C:,<

DPO-2020-002

¥,'i'-1*

(09-2019)

• n NRC MD 10,159

• 0 i

DIFFERING PROFESSIONAL OPINION Date Received

<a.,, * ...,tt** 08/05/2020 Name and Title of Submitter Organization Telephone Number (10 numeric digits)

John G Parillo NRR/DRNARCB Name and Title of Supervisor Organization Telephone Number (10 numeric digits)

Kevin Hsueh Radiation Protection and Consequence (ARCS)

Branch Chief NRR/DRNARCB (301) 415-7256 When was the prevailing staff view, existing decision or stated position established and where can it be found?

Date 07/2000 Where (i.e., ADAMS ML#, if applicable): ML003716792 Subject of DPO RG 1.183 Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors Summary of prevailing staff view, existing decision, or stated position.

See attached word document, "Crediting Safety Systems with a Deterministic Source Term."

Reason for DPO, potential impact on mission, and proposed alternatives, See attached word document, Crediting Safety Systems with a Deterministic Source Term."

Describe the (a) importance of prompt action on the Issue, (b) safety significance of the issue, and (c) the complexity of the Issue, See attached word document, "Crediting Safety Systems with a Deterministic Source Term."

Do you believe the issue represents an immediate publlc health and safety concern? 0 No Yes, (Explain In box above with importance of prompt action and safety significance,)

Is the issue directly relevant to a decision pending before the Commission? 0 No Yes, Reference Document (I.e., ADAMS ML#)

0 Informal discussions took place (Identify with whom and trme frame of discussions) Extenuating circumstances prevented informal discussions Extensfve discussions on the same issue as lncorporated into DG-1199 technfcal contact and successfve members of AADB/ARCB and Branch Chiefs (from November 2009 through present.

Proposed panel members are (in priority order):

1. Elijah Dickson 3. Jerry Dozier

2. James Shea D No names of potential panel members will be provided.

Llst of area(s) of technical expertise needed to propedy assess the Issue (e.g., electtlcal engineering, operator licensing),

Design basis accident dose consequence analyses When the process Is complete, I would llke management to determine whether public release of the DPO case file (with or without redactions} Is appropriate (Select "No If you would Uke the DPO case file to be non'1)ubllc): 0 Yes No Please note that your DPO submittal may be shared on a need-to-know basis in an effort to resolve the concern, detennine the most appropriate regulatory actions in response to the concern, and Identify key agency resources to evaluate the concern.

Signature of Submitter: John G. Parillo Digitally signed by John G. Parillo Date: 2020.07.27 10:58:54 -04'00' Signature of Co-Submitter (If any):

I Submit by E-mail: I Signature of DPO Program Manager: Gladys J. Figueroa Toledo Digitally signed by Gladys J .Figueroa Toledo Date: 2020.08.11 15:15:14-04'00' QSEDPOONLY returned

-SE DPO accepted NRC FORM 680 (09-2019) Page 3 of 3

Background:

The maximum hypothetical accident (MHA), also referred to as the maximum credible accident (MCA), is that accident whose consequences, as measured by the radiation exposure of the surrounding public, would not be exceeded by any other accident whose occurrence during the lifetime of the facility would appear to be credible. The fission product release assumed for this evaluation should be based upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events. Such accidents have generally been assumed to result in substantial meltdown of the core with subsequent release into the containment of appreciable quantities of fission products. These evaluations assume containment integrity with offsite dose consequences evaluated based on design basis containment leakage. 1 In the early stages of the development of siting criteria there were some who proffered a mechanistic cause for the MHA. As stated in TID-14844 2:

For pressurized and boiling water reactors, for example, the "maximum credible accident" has frequently been postulated as the complete loss of coolant upon complete rupture of a major pipe, with consequent expansion of the coolant as flashing steam, meltdown of the fuel and partial release of the fission product inventory to the atmosphere of the reactor building.

Through regulatory guidance the MHA has been defined as a loss of coolant accident (LOCA),

first in regulatory guide (RG) 1.3 3 and RG 1.4 4 and subsequently incorporated into RG 1.183 5 and RG 1.195 6. The definition of an MHA as a LOCA has resulted in ongoing issues resulting from the NRC staff proffering physical explanations for the use of a deterministic source term.

These issues include defining when the source term is assumed to be released into the containment atmosphere and the precise location of the assumed pipe rupture.

Discussion:

Defining the MHA as a LOCA has resulted in confusion in terms of the regulatory analysis performed for conformance with off-site and control room dose acceptance criteria and the regulatory analysis performed for conformance with § 50.46. 7 The LOCA dose consequence analysis assumes a deterministic substantial fuel melt source term while conformance with § 50.46 requires the use of mechanistic evaluations to ensure that a rupture of the largest main 1

See 10 CFR 50.34, Contents of applications; technical information, and 10 CFR 100.11, Determination of exclusion area, low population zone, and population center distance.

2 Technical Information Document (TID)-14844, Calculation of Distance Factors for Power and Test Reactor Sites, USAEC, March 23, 1962. ADAMS Accession No. ML021720780.

3 Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors, 11/1970. (Withdrawn - See 81 FR 88710, 12/08/2016) 4 Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors, 11/1970. (Withdrawn - See 81 FR 88710, 12/08/2016) 5 Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, 07/2000.

6 Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors, 05/2003.

7 10 CFR 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors

coolant piping will not result in core temperatures in excess of 2,200 ºF, well below the temperature required for fuel melt.

A proposed solution to these issues would be to make the following distinctions in guidance:

1. The design basis LOCA should be defined as the mechanistic evaluation used to show compliance with § 50.46 where the actual plant response to a major break in coolant piping is modelled to predict the resulting maximum core temperature.

And:

2. The dose consequence analysis to demonstrate compliance with predetermined dose criteria as specified in regulations 8 should be defined as the MHA.

The LOCA dose consequence analysis should be replaced with the non-mechanistic MHA.

Defining the accident as an MHA would eliminate inconsistencies as to defining its specific cause. The MHA (referred to as the LOCA in current guidance) differs from all other design basis accidents (DBAs) in that the degree of fuel damage is deterministic. In all other DBAs the licensee/applicant specifies the degree of fuel damage, if any, in its dose consequence analysis based on mechanistic analyses. For DBAs that do not result in fuel damage the source term is governed by technical specification (TS) limited coolant activity. The TS coolant activity limits can be selected by the licensee/applicant to accommodate the dose consequence analyses ensuring that the results meet predetermined acceptance criteria.

The MHA dose consequence analysis should avoid mechanistic explanations for how the bounding deterministic fuel melt source term occurs. The MHA is a hypothetical accident whereby a deterministic source term is assumed to be released into the containment atmosphere for the purpose of challenging the plant structures, systems and components (SSCs) to deal with a fuel melt source term without exceeding predetermined dose criteria.

Therefore, it is not necessary to specify the cause of the fuel melt or link the timing of the mixing of the assumed release into containment to the temporary failure of plant SSCs.

It should not be necessary to include a main coolant system pipe rupture in the MHA dose analysis. Licensees/applicants already perform mechanistic analyses for the worst case main coolant system pipe break to show compliance with 50.46. In addition, licensees/applicants mechanistically determine the degree of fuel damage, if any, resulting from the worst case main steam line break (MSLB) accident. Typically, these MSLB analyses determine little or no significant fuel damage. Therefore, it should not be necessary to include a pipe break in the MHA. For instance, BWR licensees have typically included a break in the main steam line when computing the dose consequence from main steam line isolation valve (MSIV) leakage. The incorporation of a non-mechanistic fuel melt source term in the MHA provides sufficient conservatism in the dose analysis such that assuming particular pipe breaks to add further conservatism should not be necessary to comply with the regulatory acceptance criteria.

8 These Regulations include 10 CFR Sections 100.11, 50.34, 50.67, 52.17, 52.47, 52.147 and GDC 19.

Currently dose consequences are not evaluated for the mechanistically determined fuel cladding damage that would result from the abrupt pressure loss associated with the design basis LOCA for compliance with § 50.46. Main coolant piping is required to be seismically qualified for compliance with 10 CFR 100 Appendix A. Therefore, the probability of a break in main coolant piping (defined herein as the § 50.46 design basis LOCA) is very low and would justify using the maximum dose acceptance criteria allowed by regulation if a dose consequence analysis were to be performed for this event. Given that the design basis LOCA for compliance with § 50.46 will not result in fuel melt, and that this event would have the same dose acceptance criteria as the MHA, it follows that the dose consequences from an MHA would bound the dose consequences from a § 50.46 design basis LOCA. Since the dose consequences from the MHA would bound the dose consequences from a § 50.46 design basis LOCA, it would not be necessary to produce a dose consequence analysis using the fuel cladding damage source term released into the containment associated with the § 50.46 design basis LOCA.

Evaluation of the dose consequences from an MHA would be simplified by removing mechanistic considerations of how the source term occurs. The source term for the MHA is, and always has been deterministic or non-mechanistic. Regulations do not require an explanation for how substantial fuel melt occurs, it just does. The substantial fuel melt source term is injected into the dose consequence analysis notwithstanding the operation of safety related equipment designed to preclude significant fuel failure. All other aspects of a plants response to the accident should be credited if the systems are safety related as this designation ensures reliability to perform their safety function. Safety related pumps and valves need to be credited as performing their design functions including those systems designed to deliver water to flood the core thereby mixing the deterministic fuel melt source term into the containment atmosphere. The additional measure of defense-in-depth is gained by non-mechanistically assuming that despite the operation of these systems substantial fuel melt occurs.

NRC staff proponents of the current assumption of a delayed containment mixing predicated on the temporary failure of the ECCS often cite a 1963 Advisory Committee on Reactor Safeguards (ACRS) rationale for accepting certain ESFs as substitutes for distance, but not for the ECCS reducing the fuel melt source term as described in the regulations. In a memo from Kouts to Seaborg, Report on Engineered Safeguards, dated November 18, 1964, the ACRS addressed an industry proposal seeking credit for the ECCS to arrest the progression of core melt to favor reactor siting in less-favorable site locations. 9 The memo discusses the importance of core-spray and safety-injection systems to prevent meltdown in the event of an unlikely loss of coolant accident or a major coolant leak. Kouts acknowledges these systems might not function for several reasons in the event of an accident, such as, severed lines to the reactor vessel and low water supplies and that reliance cannot be placed on systems such as these as the sole engineered safeguards in the plant. The staff has interpreted these documents to justify the ECCS to be essential for accident prevention as evidenced by § 50.46 but not to arrest the accident progression to a substantial core melt for the purposes of dose analyses.

Allowing for the mixing of the deterministic source term from the mechanical operation of ESF systems, as described in this DPO, is a separate and distinct issue from the ACRS position not 9 Herbert Kouts, Advisory Committee on Reactor Safeguards Chairman, letter to Glenn T. Seaborg, Chairman, U.S.

AEC, subject Report on Engineered Safeguards, November 18, 1964 (available through the NRC Public Document Room).

to credit the ECCS to arrest the regulatory source term. This distinction cannot be overstated.

The issue discussed in this DPO is not a source term issue but rather a mixing issue.

Impact on RG 1.183:

RG 1.183, (Reference 1), includes the assumption that there is a two hour delay in the distribution of the deterministic source term within the containment atmosphere of Mark III boiling water reactors (BWRs). This assumption is based on the time period of approximately two hours for the in-vessel fuel melt source term to fully develop as proffered in NUREG-1465 (Reference 2). For this assumption to be valid one must assume that safety related systems designed to flood the core thereby mixing the released activity into the containment atmosphere do not operate during the initial stages of the accident. The argument for the use of this assumption is that the substantial fuel melt source term could not develop if safety related systems operated as designed. However, this argument contradicts the deterministic use of the substantial fuel melt source term. The source term is given; that is, it is superimposed on the power plant to test the adequacy of the containment and other safety related systems.

Notwithstanding the deterministic imposition of the fuel melt source term on the reactor system, safety related system actuation should be credited if the system meets the proper requirements. 10 Just as containment isolation valves are credited to close and filtration systems are credited to start, so should systems designed to flood the core be credited for the mixing which would occur from their operation. The added safety margin is achieved by not crediting this core flooding with preventing core damage. Although this concept may seem counter intuitive, this is exactly how the engineered safety feature (ESF) system leakage pathway is analyzed in the design basis dose consequence analysis. The recirculation of the substantial fuel melt source term is assumed to begin at the plant specific time that the emergency core cooling system (ECCS) recirculation phase begins notwithstanding the fact that the operation of the ECCS would by design preclude a substantial core melt source term. RG 1.183, Appendix A, Section 5 includes assumptions on ESF system leakage. RG 1.183 Appendix A, RP 5.2, states that:

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 III.D.1.1 of NUREG-0737 (Ref. A-8), would require declaring such systems inoperable. The leakage should be assumed to start at the earliest time the recirculation flow occurs in these systems and end at the latest time the releases from these systems are terminated. Consideration should also be given to design leakage through valves isolating ESF recirculation systems from tanks vented to atmosphere, e.g., emergency core cooling system (ECCS) pump miniflow return to the refueling water storage tank.

For assumptions on transport in the primary containment as described in Section 3 of RG 1.183, the mixing afforded by the safety related systems designed to reflood the core is acknowledged, however, instead of crediting these systems to operate as designed the reflood and mixing has been assumed to occur at two hours post-accident. Linking the assumed time for a reflood with 10 RG 1.183 Section 5.1.2, states that: Credit may be taken for accident mitigation features that are classified as safety-related, are required to be operable by technical specifications, are powered by emergency power sources, and are either automatically actuated or, in limited cases, have actuation requirements explicitly addressed in emergency operating procedures. The single active component failure that results in the most limiting radiological consequences should be assumed.

the timing assumed in NUREG-1465 for the early in-vessel fuel melt source term to fully materialize is a departure from the concept of a deterministic source term. By making such a linkage we are attempting to provide a mechanistic justification for the use of a deterministic source term.

RG 1.183, Appendix A Regulatory Position (RP) A-3.7, states the following:

For BWRs with Mark III containments, the leakage 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 leakage 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.

This statement implies that after two hours of delay, the reflood would presumably end the core degradation corresponding to the end of the early in-vessel release phase. In addition, the actuation of core re-flood would also mix the released activity into the containment atmosphere.

Notwithstanding the fact that current research indicates that this time may be extended appreciably, there is no reason to link the initiation or the cessation of core damage to the presumed timing of safety related systems. 11 The NUREG-1465 source term should be used deterministically as has always been the case when using the TID-14844 source term as incorporated into RGs 1.3, 1.4 and 1.195. For the alternative source term (AST) described in RG 1.183, the time for the source term to fully develop was increased from the instantaneous release described in TID-14844 to approximately two hours. This time period represented a more realistic assessment of the time it would take to reach a substantial degree of core melt.

The magnitude of the early in-vessel release source term was selected to satisfy the assumption of a substantial core melt described in the regulation. 12 The use of a deterministic source term should be independent of the operation of safety related equipment. By maintaining this linkage, we would be introducing the potential risk of invalidating our safety evaluations if the time period for the early in-vessel fuel failure increases to a time substantially greater than two hours as suggested on page 41 of SAND2008-6601 (Reference 3). 13 More recently, Sandia Report SAND2011-0128, Accident Source Terms for Light-Water Nuclear Power Plants Using High-Burnup or MOX Fuel, January 2011, (Reference

5) states the following concerning the estimation of the duration of the in-vessel release phase:

11 Sandia Report SAND2011-0128, Accident Source Terms for Light-Water Nuclear Power Plants Using High-Burnup or MOX Fuel, January 2011 12

§ 50.34 Contents of applications; technical information, footnote 6: The fission product release assumed for this evaluation should be based upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events. Such accidents have generally been assumed to result in substantial meltdown of the core with subsequent release into the containment of appreciable quantities of fission products.

13 Note that, while the core water inventory is depleted after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, lower head failure is not predicted until almost 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

This modern view of the accident progression is in significant variance with the view put forth in NUREG-1465 where [lower]

head failure was predicted to occur at about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> into such an accident. This observation is evidence that the present NUREG-1465 assumptions about early in-vessel period in terms of release rate and duration are no longer consistent with current best estimate modeling results.

The In-vessel Release Phase starts when the Gap Release is completed and ends when core debris penetrates the reactor vessel and cascades into the reactor cavity where it can interact with accumulated water and structural concrete. The median durations of In-vessel Release Phase predicted for BWRs and PWRs are much longer than specified in the NUREG-1465 Source Terms for this phase of reactor accidents. The longer duration of the In-vessel Release Phase is the most profound difference among accident analyses used here and those that were the basis of the NUREG-1465 Source Term. Prolonged core degradation is not altogether surprising. Since the development of the NUREG-1465 Source Term, modeling of core degradation has been greatly improved largely by identifying and modeling of efficient mechanisms for distribution of heat from the degrading reactor fuel to the reactor coolant system especially by natural convection processes. As a consequence, degrading core material is not predicted to become as hot as rapidly as it was in calculations of reactor accidents using the Source Term Code Package [Gieseke et al., 1986] that were the basis of the NUREG-1465 Source Term.

The accident analysis described in RG 1.183, Regulatory Position A-3.7 is based on assuming that for a two hour period following the LOCA the core is left uncovered leading to fuel melt and that the only driving force to distribute fission products from the drywell into the containment is the steaming rate from the heated core. Mechanistically this sequence of events implies that there would be no AC power, no DC power and no operator actions taken for the first two hours of the accident. Realizing that the purpose of RG 1.183 is to provide guidance for the evaluation of design basis accidents, we should be able to credit the mixing that would occur from safety related systems designed to deliver water to flood the core thereby distributing fission products out of the vessel and into the containment. Under this scenario one could argue that significant fuel damage would not be expected which of course is true. However, the assumption of substantial core melt is deterministically introduced into the safety analysis to provide the additional defense in depth approach used for the evaluation of the MHA which is referred to as a LOCA in current guidance. There is no regulatory basis to arbitrarily assume that safety related equipment would not function for some predetermined period.

To maintain a consistent regulatory framework the mixing that would occur from the operation of safety related equipment should be credited at the times for which these systems are designed to achieve operability. Assuming that safety related systems fail (beyond the accepted single failure criteria) in the dose consequence analysis is inconsistent with the established regulatory practice. Regulatory Position A-3.7 in RG-1.183 should be modified to allow a licensee to defend a mixing assumption based on the operation of qualified safety related equipment. The guidance on crediting safety related systems as explained in the regulatory guide should be consistent throughout the design basis dose consequence analysis (i.e., consistent with ESF system leakage discussed in RG 1.183 Appendix A, RP 5.2). Not crediting safety related equipment in the dose analysis is inconsistent with the concept of a design basis analysis and should remain relegated only to severe accident considerations. The term severe accident is used to define an event that is both beyond the design basis of the facility and results in significant damage to the reactor core. 14 Assuming the complete failure of safety related systems to provide mixing of the deterministic fuel melt accident source term is beyond the accepted practice for the evaluation of design basis accidents.

14 Severe Accident Management Guidance (SAMG) Training for NRC, 2010 Westinghouse Electric Company LLC.

There is no mechanistic explanation provided for why the mixing provided by safety related systems is assumed not to occur or why the restoration of such equipment is assumed to occur at approximately two hours post-accident. Assuming a complete station blackout of both AC and DC sources coupled with no operator actions for two hours would be one of the only conditions that could explain the loss of mixing provided by safety related equipment. Such an assumption is clearly outside of the bounds of a design basis accident evaluation. In addition, there is no equivalent assumption of a delay in the operation of safety related equipment made for evaluations using the TID-14844 source term. For licensees using the old source term described in RG 1.195 the release as well as mixing in the containment is assumed to be instantaneous. This difference represents an inconsistency in our regulatory guidance.

Attempting to introduce a mechanistic explanation for the use of a deterministic degree of fuel damage will result in inconsistencies throughout the dose consequence analysis. As previously described, one such inconsistency relates to the assumptions regarding the ESF leakage component to the MHA dose consequence analysis (see RG 1.183 Appendix A, RP 5.2).

Current guidance instructs licensees to assess the dose contribution from ESF leakage at plant specific time for the initiation of recirculation. This timing is and should be independent of which source term is used in the analysis. However, if the complete failure of safety related equipment is assumed for the first two hours using AST guidance, then for consistency the plant specific time for recirculation would need to be increased by two hours. In addition, this time would need to be subsequently adjusted as new insights regarding the timing of fuel damage are incorporated into regulatory guidance.

In the review of license amendment requests (LARs) to adopt the AST, licensees have been instructed, via the request for addition information (RAI) process, to delay the mixing of the drywell and suppression pool free air volumes in BWRs until two hours after accident initiation.

This delayed reflood is attributed to the timing of the AST described in RG 1.183 to justify the core melt source term. Instead of crediting safety related systems to mix the drywell free air volume with the suppression free air volume based on plant response times, licensees have been instructed to delay the mixing for two hours resulting in higher initial drywell concentrations and hence higher calculated dose consequences. BWR licensees should be allowed to revise their dose consequence analyses to remove this unnecessary delay in the operation of safety related systems and thereby gain some additional margin to the acceptance criteria.

The application of this delayed mixing assumption in DG-1199 15 (Reference 4) will have a significant impact on the ability of BWR licensees to meet regulatory dose acceptance criteria which includes the contribution from main steamline isolation valve (MSIV) leakage. Based on a two hour delay in the mixing of the release from the core into the containment, DG-1199 includes the assumption that the initial source of MSIV leakage is the reactor vessel steam dome and not the containment atmosphere. This assumption results in an increase in the calculated concentration of the initial MSIV leakage releases which leads to a significant increase in the calculated dose consequences. This is the most significant issue resulting from the current NRC staff position of not crediting safety related systems to operate at the times specified in the plants licensing basis. Since DG-1199 is not an official document the DPO 15 Draft Regulatory Guide DG-1199, (Proposed Revision 1 of Regulatory Guide 1.183, Dated July 2000)

Alternative Radiological Source Terms For Evaluating Design Basis Accidents At Nuclear Power Reactors, October 2009, ML090960464.

process cannot be used to directly challenge this assumption in DG-1199. On careful examination, the origin of this assumption can be traced back to the original RG 1.183 and therefor this DPO now addresses this assumption in an official NRC document.

==

Conclusion:==

In summary, the concept that credit for the operation of safety related systems can be granted selectively to justify the use of an AST will produce inconsistencies in the dose consequence analysis and therefore result in regulatory uncertainty. Further and more importantly, selectively crediting the mixing provided by safety related equipment has set a precedent for a design basis accident analysis which has previously been relegated to severe accident analysis thereby creating an unnecessary regulatory burden without a significant increase in safety. Finally, not crediting ESF systems to distribute the assumed fuel melt accident source term into the containment conflicts with 10 CFR 50.34 Contents of applications; technical information, § (a)(ii)(D) which states that, "an applicant shall assume a fission product release from the core into the containment."

The issues described herein can all be resolved by reinstating the MHA as the unexplained release of a deterministic fuel melt source term into an intact containment atmosphere. With this definition there is no need to explain how the source term occurs or to specify how and when the source term is mixed into the containment atmosphere. This solution will simplify the calculation of the MHA dose consequences and relieve unnecessary regulatory burden.

References:

1. Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000

2. NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants, February 1995.

3. Gauntt, R.O. 2008, Analysis of Main Steam Isolation Valve Leakage in Design Basis Accidents Using MELCOR 1.8.6 and RADTRAD, SAND2008-6601, Sandia National Laboratories, Albuquerque, NM.

4. Draft Regulatory Guide DG-1199, (Proposed Revision 1 of Regulatory Guide 1.183, dated July 2000), Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, October 2009.

5. Sandia Report SAND2011-0128, Accident Source Terms for Light-Water Nuclear Power Plants Using High-Burnup or MOX Fuel, January 2011

From: Parillo, John To: Benton, Laray; Blumberg, Mark; Boatright, Aleem; Brown, Leta; Duvigneaud, Dylanne; Parillo, John; Rautzen, William; Shea, James; Tate, Travis; White, Jason Cc: Taylor, Robert

Subject:

Crediting safety system operation with a deterministic source term - Sreela Ferguson"s comment Date: Wednesday, November 18, 2009 2:22:00 PM Colleagues, The purpose of the footnoted source term descriptions in both Part 100 and 50.67/50.34 is to provide a maximum credible release of radionuclides to test the plant safety systems against accident dose limits. With the TID instantaneous release source term, the subsequent mixing of the source term into the containment atmosphere was never questioned. With the delayed alternative source term the question of when effective mixing takes place has been clouded by the selection of an unmitigated source term up to the end of the in-vessel release phase. NUREG-1465 estimates that this will occur at approximately two hours. My sense is that the Commission decided that truncation of the source term degradation at the end on the in-vessel release phase would provide the substantial amount of core damage required to satisfy the intent of the regulation.

It is not clear to me that the use of the in-vessel release source term should necessarily translate into the mechanistic assumption the ECCS failed to operate for the first two hours of the accident but rather that the ECCS, while mechanically operating, fails to effectively prevent substantial core damage from occurring. This is what is done when evaluating the dose consequence from the ECCS leakage pathway; the mechanical action of the ECCS is acknowledged while the core quenching aspects of the ECCS are ignored.

The suggestion by Sreela Ferguson in Monday's public meeting was that for the MSIV leakage pathway the same assumption could be used; namely that the mixing action caused by ECCS operation could be credited while ignoring the core quenching aspects of the ECCS. In essence the deterministic source term should be superimposed on safety systems that are assumed to be operating as designed. The defense in depth is covered by deterministically forcing a degree of core damage far in excess of that which would be expected to occur. It is not consistent with the treatment of other safety systems to artificially negate the mechanical operation of pumps, valves, etc. for two hours or for any arbitrary period. Moreover, current state of the art core damage assessment suggests that substantial core damage is unlikely for periods significantly greater than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. If NUREG-1465 is updated it is likely that the end of the in-vessel release phase would be substantially increased beyond the current estimated two hour value.

I would suggest we give some serious thought to Sreela's comment by considering the systematic decoupling of the mechanistic action of safety related equipment from the deterministically imposed maximum credible source term.

As always, your feedback will be greatly appreciated.

+++++++++++++++++++++++++++++++

John G. Parillo Reactor Engineer US Nuclear Regulatory Commission Office of Nuclear Reactor Regulation

Division of Risk Assessment Accident Dose Branch Office: O-10H18 / (301) 415-1344 Mail Stop: O-10C15

+++++++++++++++++++++++++++++++

The contents of this message are mine personally and do not necessarily reflect any position of the NRC.

Chronology:

14APR2009: Non-Concurrence (NCP) on DG-1199. This NCP included several issues in the modeling of MSIV leakage including challenging the assumption of the steam dome being the initial source of MSIV leakage for the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> prior to the assumed mixing into the containment atmosphere.

11JUN2009: Response to Non-Concurrence on Draft Regulatory Guide DG-1199, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors. The conclusion stated that the issues raised in the NCP would be revisited after the receipt of public comments.

16NOV2009: Public meeting on DG-1199 where a commenter raised the issue of crediting safety related systems to mix the deterministic source term at the times defined in the plants licensing basis; the source term is given.

18NOV2009: E-Mail reinforcing comment from public meeting that credit should be given for safety related systems to mix the deterministic source term at the times defined in the plants licensing basis.

02JAN2014: After numerous discussions with dose assessment reviewers, the technical contact and successive branch chiefs with no resolution, an attempt was made to file a Non-Concurrence on DG-1199.

The NCP was not accepted because the document was not in concurrence.

10MAR2019: After numerous discussions with dose assessment reviewers, the technical contact and successive branch chiefs yielding no resolution, another attempt was made to file a Non-Concurrence on DG-1199. Again, the NCP was not accepted because the document was not in concurrence.

03MAR2019: A DPO was submitted on DG-1199. Office of enforcement Staff stated that they were not certain that a DPO can be submitted on a draft RG.

01APR2019: Discussion was held with DRA management on the submitted DPO. Management expressed their preference to handle this issue through the NCP than through the DPO process. It was agreed that the DPO would be withdrawn with the understanding that DRA management intended to finalize RG 1.183 within the current calendar year.

16SEP2019: In a meeting to discuss the path forward for RG 1.183 it was decided that the proposed revision to RG 1.183 will be put on hold for approximately two years while several license actions involving main steam isolation valve leakage are completed.

04OCT2019: The previously submitted DPO on DG-1199 was re-submitted due to the continued delay in addressing the issues raised by the author.

18NOV2019: A meeting was held to facilitate communication between the submitter and the decisionmaker regarding the DPO submitted on DG-1199 to try to resolve it and/or identify a path forward. The representatives from DPOPM Resource informed all present that since the DPO was submitted on a draft RG the position of the DPOPM is that the DPO could not be accepted for formal resolution since a draft RG is not an official NRC document.

31JUL2020: After careful examination of the original RG 1.183 (July 2000) the same issue presented in the previously submitted DPO on DG-1199 (a 2-hour delay in mixing the source term into the containment) was found in the current RG 1.183. Therefor to achieve a timely closure of the issue previously raised as early as

18NOV2009, the current DPO on RG 1.183 was written to capture the same issue as incorporated in an official NRC document. It is hoped that submittal of this revised DPO will generate a resolution to this issue.

Document 2: Memo Establishing DPO Panel UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 August 24,2020 MEMORANDUM TO: Kevin Coyne, Panel Chairperson Office of Nuclear Material Safety and Safeguards Michelle Hart,Panel Member Office of Nuclear Reactor Regulation Elijah Dickson,Panel Member Office of Nuclear Reactor Regulation George A. Digitallysigned by George THRU: George A. Wilson,Director A. Wilson Office of Enforcement Wilson Date; 2020.08.24 04:19:04

-04'00' FROM: Gladys J. Figueroa-Toledo IRA/

Differing Views Program Manager Office of Enforcement

SUBJECT:

AD HOC REVIEW PANEL - DIFFERING PROFESSIONAL OPINION ASSOCIATED WITH THE RG 1.183, ALTERNATIVE RADIOLOGICAL SOURCE TERMS FOR EVALUATING DESIGN BASIS ACCIDENTS AT NUCLEAR POWER REACTORS (DPO-2020-002)

In accordance with Management Directive (MD) 10.159, "The NRC Differing Professional Opinion Program/ and in my capacity as the Differing Professional Opinion (DPO) Program Manager; and in coordination with George Wilson,Director, Office of Enforcement, Raymond Furstenau, Director, Office of Nuclear Regulatory Research; and the DPO submitter; you are being appointed as members of a DPO Ad Hoc Review Panel (DPO Panel) to review a DPO submitted by an U.S. Nuclear Regulatory Commission (NRC) employee.

The DPO (Enclosure 1) involves the RG 1.183,Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors. The DPO has been forwarded to Mr. Furstenau for consideration and issuance of a DPO Decision.

CONTACT: Gladys Figueroa-Toledo,OE (301) 287-9497 Ian Gifford,OE (301) 287-9216

K. Coyne, et al. 2 The DPO Panel has a critical role in the success of the DPO Program. Your responsibilities for conducting the independent review and documenting your conclusions in a report are addressed in the handbook for MD 10.159 in Section II.F and Section II.G, respectively. The DPO Web site also includes helpful information, such as a Differing Views Best Practices Guide, tables with status information and timeliness goals for open DPO cases , and closed DPO case files (which include DPO panel reports). We will also be sending you additional information that should help you implement the DPO process.

Timeliness is an important DPO Program objective. Thus, the disposition of this DPO should be considered an important and time sensitive activity. Although the DPO MD identifies a timeliness goal of 75 calendar days for the DPO panel review and report and 21 additional calendar days for the issuance of a DPO Decision, the DPO Program also sets out to ensure that issues receive a thorough and independent review. Therefore, the overall timeliness goal will be based on the significance and complexity of the issues, schedule challenges, and the priority of other agency work. Process Milestones and Timeliness Goals specific to this DPO will be discussed and established at a kick-off meeting.

Communication of expected timelines and status updates are important in the effectiveness and their overall satisfaction with the Differing Views Program. If you determine that your activity will result in the need for an extension beyond your timeliness goal, please send an e-mail to Mr.

Furstenau, the DPO submitter, and DPOPM.Resource@nrc.gov and include the reason for the extension request and a proposed completion date for your work. Mr. Furstenau is responsible for subsequently forwarding the request for a new DPO Decision issuance timeliness goal to the EDO for approval.

An important aspect of our organizational culture includes maintaining an environment that encourages, supports, and respects differing views. As such, you should exercise discretion and treat this matter appropriately. Documents should be distributed on an as-needed basis.

In an effort to preserve privacy, minimize the effect on the work unit, and keep the focus on the issues, you should simply refer to the employee as the DPO submitter. Avoid conversations and refrain from behaviors that could be perceived as retaliatory or chilling to the DPO submitter or that could potentially create a chilled environment for others. It is appropriate for employees to discuss the details of the DPO with their co-workers as part of the evaluation; however, as with other predecisional processes, employees should not discuss details of the DPO outside the agency. If you have observed inappropriate behaviors, heard allegations of retaliation or harassment, or receive outside inquiries or requests for information, please notify me or Ian Gifford.

On an administrative note, please ensure that all DPO-related activities are charged to Activity Code ZG0007. Managers should report time to their Management/Supervisor Activity Code. Administrative Assistants should report time to their Secretary/Clerical Activity Code.

We appreciate your willingness to serve and your dedication to completing a thorough and objective review of this DPO. Successful resolution of the issues is important for NRC and its stakeholders. If you have any questions or concerns, please feel free to contact me or Ian Gifford. We look forward to receiving your independent review results and recommendations.

K. Coyne, et al. 3

Enclosures:

1. DPO-2020-002 Submittal

2. Process Milestones and Timeliness Goals cc: R. Furstenau, RES S. Coffin, RES N. Difrancesco, RES J. Tappert, NMSS M. Hayes, NRR K. Hsueh, NRR J. Parillo, NRR G. Wilson, OE F. Peduzzi, OE D. Solorio, OE G. Figueroa-Toledo, OE I. Gifford, OE

K. Coyne, et al.

SUBJECT:

AD HOC REVIEW PANEL - DIFFERING PROFESSIONAL OPINION ASSOCIATED WITH THE RG 1.183, ALTERNATIVE RADIOLOGICAL SOURCE TERMS FOR EVALUATING DESIGN BASIS ACCIDENTS AT NUCLEAR POWER REACTORS (DPO-2020-002) DATE:08/21/2020 ADAMS Package: ML20234A537 MEMO: ML20234A555 - ML20225A165 - ML20234A566 OE-011 OFFICE OE: DPO/PM OE:D NAME GFigueroaToledo GWilson DATE 08/20/2020 08/24/2020 OFFICIAL RECORD COPY

Document 3: DPO Panel Report UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O,C, 20555-0001 January 19, 2021 MEMORANDUM TO: Ray Furstenau, Director Office of Nuclear Regulatory Research FROM: Kevin Coyne, Panel Chairperson IRA/

Office of Nuclear Material Safety and Safeguards Michelle Hart, Panel Member IRA/

Office of Nuclear Reactor Regulation Elijah Dickson, Panel Member IRA/

Office of Nuclear Reactor Regulation

SUBJECT:

DIFFERING PROFESSIONAL OPINION ASSOCIATED WITH THE REGULATORY GUIDE 1.183, ALTERNATIVE RADIOLOGICAL SOURCE TERMS FOR EVALUATING DESIGN BASIS ACCIDENTS AT NUCLEAR POWER REACTORS (DPO-2020-002)

In a memorandum dated August 24, 2020, we were appointed as members of a Differing Professional Opinion (DPO) Ad Hoc Review Panel (DPO Panel or Panel) to review a DPO regarding the Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors. The DPO Panel has reviewed the DPO in accordance with the guidance in Management Directive 10.159, "The NRC Differing Professional Opinion Program." The scope was limited to a review of the issues identified in the DPO as clarified through a Summary of Issues developed by the Panel and confirmed by the DPO Submitter. The Panel evaluated the issues through interviews of knowledgeable NRC staff and a review of various documents, including official agency records.

The results of the DPO Panel's evaluation of the concerns raised in the DPO are detailed in the enclosed DPO Panel Report. Based on our review of concerns raised in the DPO, the DPO Panel made the following conclusions:

• Current regulatory practices for design-basis offsite consequence analysis have included certain mechanistic system performance assumptions (consistent with a core damage source term) in order to maximize offsite radiological consequences. Although these assumptions are not specifically required by regulations, they have provided additional margin to account for severe-accident uncertainties and were intended to ensure the protection of public health and safety.

CONTACT: Elijah Dickson, NRR/DRR 301-415-7647

Use of the NUREG-1465 Source Term at Operating Reactors

ML21015A404 NRR-106 OFFICE NRR/DRA/ACRB NRR/DANU/UART NMSS/REFS NAME EDickson MHart KCovne DATE 1/15/2021 1/19/2021 1/17/2021 DIFFERING PROFESSIONAL OPINION ON THE NRC ASSUMPTION OF CREDITING SAFETY SYSTEMS WITH A DETERMINISTIC SOURCE TERM AND THE MAXIMUM HYPOTHETICAL ACCIDENT (DPO-2020-002)

Differing Professional Opinion Panel Report IRA/

Kevin Coyne, Panel Chair IRA/

Michelle Hart, Panel Member IRA/

Elijah Dickson, Panel Member January 22, 2021 Date Enclosure

Code of Federal Regulations Accident Source Term,

leak rate from the containment and the meteorological conditions pertinent to his site."

Each of these regulations reference footnotes which define the source term that should be assumed for these evaluations. For example, the footnote referenced by 10 CFR 50.34(a)(1)(ii)(D) states:

The fission product release assumed for this evaluation should be based upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events. Such accidents have generally been assumed to result in substantial meltdown of the core with subsequent release into the containment of appreciable quantities of fission products.

Although these regulations reference the use of a source term involving substantial meltdown of the core, they do not provide prescriptive requirements or require licensees to use specific designs or methodologies to comply with the regulations. Furthermore, although the regulations include consideration of plant systems, structures, and components (SSCs) and site characteristics, they do not provide prescriptive requirements on how the transport of fission products from the core to the environment should be modeled. Instead, the NRC's regulatory guidance (e.g. RG 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors") provides an acceptable source term and methods and assumptions for modeling the plant response and transport of radionuclides to the environment and calculation of radiological consequences for demonstration of compliance with the radiological performance criteria. Therefore, the important distinction regarding the issue is not how the source term is derived and defined, but rather how the source term transport to the environment is modeled. Furthermore, the assumptions regarding radionuclide transport to the environment are driven by whether or not SSCs should be modeled in a way that is consistent with a substantial meltdown of the core or if the operation of these SSCs should confirm to normal design basis assumptions. This dichotomy is illustrated by the current differing perspectives among the staff:

• Mechanistic Perspective on Source Term:

Staff aligned with a mechanistic perspective on the source term attempt to model SSC functionality during radionuclide transport in a manner that is consistent with the generation of a substantial meltdown of the core. For example, the source term described in NUREG-1465 (USNRC, 1995) is derived from mechanistic considerations and is a major improvement over the traditional TID-14844 (USAEC, 1962) deterministic source term. The NUREG-1465 accident source term is more physically-based using probabilistic risk assessment, post Three Mile Island analyses, and experiments to support the technical bases of its development. The mechanistic approach couples the thermal-hydraulic behavior of the degrading reactor core through the primary coolant system to the postulated breach into the containment as a precondition for the calculation of the radionuclide transport.

Several severe-accident scenarios were used to develop the NUREG-1465 accident source term, including scenarios with a non-functioning emergency core cooling system (ECCS), which leads to earliest onset of core damage. Therefore, staff applying this perspective reason that in order to assess radionuclide transport, including containment performance and mitigating engineered safety features (ESFs), one should assume operation of these safety systems matches, to the extent

always transport transport

Panel about the way the issues were stated. As a result, the final list of issues was finalized on October 7, 2020.

2.1 As-Interpreted by the DPO Panel Based on Initial DPO Submittal The DPO Submitter's fundamental disagreement is with certain aspects of the regulatory guidance and staff practices associated with the assessment of ESFs when applied to mitigate or mix the radiological consequences of nuclear power plant accidents. These evaluations are conducted pursuant to several regulatory requirements, including; 10 CFR 50.67, "Accident source term," 10 CFR 50.34(a)(1)(ii)(D), 10 CFR 50, Appendix A, General Design Criterion 19, "Control Room," and 10 CFR 100.11, "Determination of the exclusion area, low population zone, and population center distance." In accordance with various footnotes found throughout the regulations1 , the fission product release assumed for these evaluations should be based upon a major accident, hypothesized for the purposes of site analysis or postulated from considerations of possible accidental events" and which have generally been assumed to result in "substantial meltdown of the core. RG 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," provides guidance for conducting these radiological assessments for licensees implementing an alternate source term under 10 CFR 50.67.2 The NRC's regulatory guidance documents specify an assumed fission product release from the core that conforms to the footnote language that the fission product release is intended to not be exceeded by any accident considered to be credible. The fission product release associated with this evaluation is referred to as a "maximum hypothetical accident3" (MHA) source term in this DPO report. As noted in RG 1.183, DBAs were not intended to be actual event sequences, but rather, were intended to be surrogates to enable deterministic evaluation of the response of a facility's ESFs. To this end, RG 1.183 provides a deterministic source term for the loss-of-coolant accident (LOCA) which has been used for the evaluation of an MHA. For non-LOCA events, RG 1.183 provides estimated gap fractions for radiological consequence-significant radionuclides but the degree of fuel cladding damage and fuel centerline melt, if any, is determined from site-specific mechanistic evaluations.

The NRC also evaluates a spectrum of design-basis events to ensure that the capabilities of safety-related SSCs provide certain safety functions, including maintaining the integrity of the reactor coolant pressure boundary, safe shutdown, and mitigation of offsite exposures4* The spectrum of design-basis events considered in these evaluations includes anticipated operational events and postulated accidents. In addition, 10 CFR 50.46, "Acceptance criteria 1 For example, see (1) 10 CFR 100.11 "Determination of exclusion area, low population zone and population center distance," Footnote 1, (2) 10 CFR 50.67, "Accident source term," Footnote 1; (3) 10 CFR 50.34, "Contents of applications; technical information," Footnotes 6 and 11 for the PSAR and TMI requirements respectively. For new reactor applications pursuant to Part 52, there are similar requirements, including the footnote.

2 RG 1.195, Methods and Assumptions for Evaluating Radiological Habitability Assessments at Nuclear Power Plants" provides similar guidance for licensees using the original source term specified in TID-14844.

-J This event has also been referred to as ac "maximum credible accident'. RG 1.183 uses the term "design basis accident to refer to the sour e term associated with these types of fission product release events.

4 As noted in 10 CFR 50.2 "Definitions," and 10 CFR 50.49 "Environmental qualification of electric 1 .1 equipment important to sarety for nuclear power plants," sarety-related structures, systems and components are relied upon to remain functional during and following design basis events to ensure that certain safety requirements are met.

event analyses (where significant fuel damage is precluded). In doing so, the staff would be consistent with crediting certain aspects of safety-related systems by superimposing the deterministic source tenn on these safety systems that are designed to be operating.

2.1.1 Statement of Issues

1. Meeting intent of regulation at hand - The regulation requires an analysis of a large release to an intact containment and does not specify an accident scenario, but RG 1.183 describes analysis assumptions for a LOCA which leads to: (1) an inappropriate need to impose non-physical assumptions pertaining to the operation of safety-related SSCs to mechanistically account for the DBA source term; (2) inconsistency with LOCA analysis for 10 CFR 50.46; and (3) over-conservative assumptions intended to maximize fission product release (e.g., RG 1.183 assumptions on Mark Ill drywell to containment mixing).

2. Process issue - Ongoing licensing issues resulting from the staff proffering physical explanations for the deterministic source term. For example, licensees have been instructed, via the request for additional information (RAI) process, to delay the mixing of the drywell and suppression pool free air volumes in any type of BWR (not just those with Mark Ill containments, as noted in RG 1.183, Appendix A, Assumption 3. 7) until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after accident initiation with respect to the containment leakage pathway.

3. Crediting safety-related SSCs in analyses - Inconsistent treatment of crediting safety related SSCs within the deterministic radiological consequence analyses when determining compliance with the various DBA radiological performance criteria.

3. EVALUATION 3.1 Background Discussion Before presenting the Panel's assessment of the differing professional option, briefs are provided on three pertinent topics to provide the necessary background:

1. Regulatory requirements and source terms;

2. General staff practices of assessing DBA radiological consequences; and,

3. Risk-informed decision-making information.

3.1.1 Regulatory requirements and source terms Use of regulatory source terms in design-basis accident radiological consequence assessments is deeply embedded in the regulatory policy and practices of the NRC, even as the licensing process has evolved over the past 50 years. As defined in 10 CFR 50.2, the source term refers to the magnitude and mix of the radionuclides released from the fuel, expressed as fractions of the fission product inventory in the fuel, as well as their physical and chemical form, and the timing of their release. A source term based upon substantial meltdown of the core supports the concept of defense-in-depth in which power plant design, operation, siting, and emergency planning comprise independent layers of nuclear safety. This approach encourages nuclear plant designers to incorporate several lines of defense in order to maintain the effectiveness of physical barriers between radiation sources and materials from workers, members of the public and environment in operational states and, for some barriers, in accident conditions. Within this

Reactor Site Criteria Calculation of Distance Factors for Power and Test Reactors An Assessment for Five U.S. Nuclear Power Plants (NUREG-1150)

Accident Source Terms for Light-Water Nuclear

Power Plants (NUREG-1465). The NRC's intent in publishing NUREG-1465 was to capture the major relevant insights available from severe-accident research to provide, for regulatory purposes, a more realistic portrayal of the amount of the postulated accident source term6 These revised accident source terms are more realistic, particularly in the areas of timing in the release of radioactivity to containment and chemical classes of radionuclides. To account for elements that have the potential of producing significant consequences, the NUREG-1465 source term defines eight chemical classes of radionuclides rather than the three physical classes considered in the TID-14844 source term as well as highlights the importance of radionuclides released to the containment as aerosol particles. Also, timing of the phases and release fractions are taken to be different for PWRs and BWRs.

In SRM-SECY-96-242, "Use of the NUREG-1465 Source Term at Operating Reactors,"

(USNRC, 1996) the Commission approved the staffs approach to allow use of the revised accident source terms and commence rulemaking upon completion of the new source term rebaselining activities concurrent with the pilot plant evaluations. The Commission provided direction to the staff to clarify when developing guidance, "[t]he staff should exercise caution so as to avoid creating new severe-accident mitigation requirements in the licensing of currently operating plants that do not have explicit and informed Commission approval."

In December 1999, the NRC issued a new regulation, 10 CFR 50.67, Accident source term, which provided a mechanism for licensed power reactors to voluntarily replace the traditional TID-14844 accident source term used in their design-basis accident analyses with an alternative source term more consistent with the results published in NUREG-1150 and NUREG-1465. As noted in the statements of consideration for 10 CFR 50.67, the NRC determined that design basis analyses will address the first three release phases in NUREG-1465 --coolant, gap, and in-vessel. The ex-vessel and late in-vessel phases are considered to be inappropriate for design-basis analysis purposes. These latter releases could only result from core damage accidents with vessel failure and core-concrete interactions. (See also SECY-94-302, "Source Term-Related Technical and Licensing Issues Pertaining to Evolutionary and Passive Light Water-Reactor Designs" (USNRC, 1994)). Regulatory guidance for the implementation of the AST is provided in RG 1. 183, which adapts the revised accident source terms provided in NUREG-1465 for licensing analyses. RG 1.183 describes a maximum credible accident with a non-specific initiator that is conceptually an arrested core damage accident that stops the analysis at the early in-vessel release phase. RG 1.183 also provides guidance for determining the relative distribution of the source term between the reactor vessel and the various containment volumes (e.g., between the drywell suppression pool for a BWR). While these source term distribution assumptions are generally consistent with the accident sequences considered in developing NUREG-1465 (e.g., delayed actuation of ECCS), they may not be consistent with expected design-basis operation of ESFs. However, RG 1.183 does state that credit may be taken for accident mitigation features that meet certain criteria (e.g., safety related, required by technical specifications, powered by emergency power supply). To date, nearly all commercial nuclear power plant licensees have adopted the 10 CFR 50.67 AST as their licensing and design-basis by applying the methodologies of RG 1.183.

The promulgation of 10 CFR 50.67 and provision of the methodologies described in RG 1.183 created an inflection point in how the NRC staff made a safety finding regarding the approval of changes at nuclear power plants when computing OBA radiological consequences. It has been recognized since the development of the revised source term that changes in the prescription of 6

See the Statements of Consideration for 10 CFR 50.67 (64 FR 72001)

Management of Backfitting, Forward Fitting, Issue Finality, and design-basis both explicitly and implicitly the NUREG-1465 source term. For instance, the explicit timing-duration and release fractions and the implicit assumption that the ECCS functionality is delayed for the first two hours of the accident to meet the source term distribution guidance provided in RG 1.1837*

Appendix A provides a broad assessment of staff safety evaluation reports approving 10 CFR 50.67 AST license amendments regarding the assumption of drywell/wetwell mixing after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to effectively match the NUREG-1465 source term.

In 2011, Sandia National Laboratory published, Accident Source Terms for Light water Nuclear Power Plants Using High-Burnup or MOX Fuel, (SAND2011-0128)

(Powers, 2001), on the behalf of the NRC, which provided representative accident source terms patterned after NUREG-1465 for high-burnup fuel using an updated version of the MELCOR severe-accident analysis computer code.

This new source term research identified several important differences compared to NUREG-1465. For example, SAND2011-0128 predicts a significantly longer time-duration of the early in-vessel release period compared to NUREG-1465.

Specifically, in SAND2011-0128, the early in-vessel release extends to 4.5- and 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for PWRs and BWRs, respectively, instead of the 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> given in NUREG-1465. Release group fractions are comparable for most chemical classes, but there are some differences, the most notable being a higher iodine release fraction for BWRs in the 2011 study. The proposed source terms from SAND2011-0128 are technically justifiable and meet the attributes of an acceptable source term as defined in Section C.2 of RG 1.183. However, the differences between the two source term vintages is not necessarily due to an improved understanding of severe-accident phenomena. Rather, substantial improvements in the computational modeling capabilities (e.g. increased computing power and the ability to nodalize the model to produce for finer resolution results) has lead to the better understanding of reactor core accident meltdown physics and subsequent prediction of radionuclide release and behavior.

Clearly, over time improved modeling capabilities and new science leads to a better understanding of severe-accident progression, as demonstrated by the SAND2011-0128 study. More recent research studies, such as the State of the Art Consequence Analysis (NUREG-1935), further demonstrate this trend towards more realistic analysis showing that severe accidents take much longer to happen and release much less radioactive material than earlier analyses suggested. (USNRC, 2012). This circumstance leads to potential regulatory instability, as source term characteristics may change as our knowledge improves. However, it is not clear how, or if, this knowledge would be applied for licensing purposes. As discussed above, and highlighted in the DPO, every US nuclear power plant regulated by the NRC which has adopted 10 CFR 50.67 has updated their licensing- and design bases to incorporate the NUREG-1465 gap and early in-vessel phase results and describe a scenario consistent with the 7

The delay of ESFs appears to run counter to RG 1.183, Regulatory Posttion 5.1.2, which permits credit for accident mitigation features that are classified as safety-related, are required to be operable by technical specifications, are powered by emergency power sources, and are either automatically actuated or, in limited cases, have actuation requirements explicitly addressed in emergency operating procedures.

source term assumptions by mechanistically assuming no ECCS performance for the first two hours of the accident. Were the NRC to adopt the SAND2011-0128 study results within an updated version of RG 1.183 (which is currently being revised at the same time this DPO is being considered) and still maintain the mechanistic linkage to not credit the mixing effects. of safety-related ESF functionality before the end of the extended early in-vessel phase, the concern is that the agency could be introducing the potential risk of invalidating previous safety evaluations based on the previous 2-hour assumption. This would also call into question the issues of backfitting, forward fitting, and issue of finality.

By delaying ECCS operation in the radionuclide transport modeling, the design related dose values associated with this release phase would simply increase; possibly above the regulatory radiological performance criteria which would create a compliance issue with the applicable regulations. This is due in part to the stylized nature of the radiological consequence analyses. However, there appears to be no substantial increased safety benefit by arbitrarily mechanistically matching ECCS performance to the core damage release phase timing of the SAND2008-0128 study results for the purposes of evaluating fission product transport after its release from the core. Furthermore, this approach can improperly link a design-basis analysis to SSC performance assumptions that are beyond the plant design and licensing basis. This lends credence to the DPO's thesis that for the purposes of challenging the plant safety-related ESF SSCs to deal with a core melt source term without exceeding predetermined radiological performance criteria, the staff should refrain from imposing mechanistic explanations for how the accident occurs by disregarding the availability of certain safety-related SSCs which are expected to operate as designed when to credit their effects on transport and mixing of fission products.

EXAMPLE 2: RG 1.183 does not provide specific guidance for aerosol deposition within the main steam lines of a BWR. Instead, licensees have followed the initial analyses supporting the first approved 10 CFR 50.67 AST implementation for the Perry plant, Assessment of Radiological Consequences for the Perry Pilot Plant Application using the Revised (NUREG-1465) Source Term, (AEB-98-03).8 The method was found acceptable for use with certain conditions to address issues as discussed in Regulatory Issue Summary (RIS) 2006-04, NRG Regulatory Issue Summary 2006-04, Experience with Implementation of Alternate Source Terms. (USNRC, 2006). Subsequently, this aerosol de.position model, as approved, is now part of the licensing bases for most operating BWRs. Since the publication of AEB 9S-03, the parameters for estimating the aerosol velocity distribution and use of a single median settling velocity have come into question.

Specifically, this approach does not account for the quicker settling and removal of heavier and larger aerosol particles compared to lighter and smaller particles.

These issues are discussed in further detail in NRC RIS 2006-04.

In 2008, Gaunt, et al. (2008), published the report entitled, Analysis of Main Steam Isolation Valve Leakage in Design Basis Accidents Using MELCOR 1.8.6 and RAD TRAD, (SAND2008-6601) on the behalf of the NRC. The purpose of this report was to assess the reactor accident source terms for the BWR main 8

ADAMS Accession Number ML011230531

steam line isolation valve leakage pathway, and radioactive aerosol behavior.

The approach was to develop a model to replace the current practice of using the containment AST source term (NUREG-1465) airborne aerosol concentrations, required by regulations, as a surrogate for the in-vessel aerosol concentration that exists within the reactor steam dome. This steam dome concentration would then be used to compute releases to the environment from the MSIVs leakage rates specified in the technical specifications9* The intent was to correct the previous issues identified in RIS 2006-04 and to provide a uniform model for licensees to follow in support of the NRC's reviews of license amendment requests implementing the AST pursuant to 10 CFR 50.67. Part of this formulae was to characterize the vessel steam dome source term through the use of scaling factors as a transport model to scale the drywell airborne concentration determined from the stylized AST (containment source) to a value for the vessel steam dome. SAND2008-6601 discusses a slight complication to this approach since the time NUREG-1465 was issued, the then-current MELCOR best estimate predictions on the timing and magnitude of releases from the core to the containment had changed from that described in the NUREG-1465 prescription (Example 1 describes this). In particular, releases to the containment are now found to be delayed in time and to occur at a lower rate. So, in order to account for this difference between the then-current MELCOR predictions of containment airborne concentrations and those determined by the NUREG-1465 methodology, an additional correction was necessary to "normalize" the results.

This normalization is the ratio of the then-current MELCOR-predicted containment airborne concentrations to the NUREG-1465 predicted containment airborne concentrations, referred to as "R*-factors". As described in the report, this normalization is necessary because NUREG-1465 containment airborne concentrations, when scaled up by MELCOR steam dome to containment concentration ratios, produced an excessively conservative result.

Analyses showed that if the SAND2008-6601 model were applied, computed radiological consequences would increase over the current licensing basis analysis results and could exceed the applicable regulatory radiological performance criteria creating a possible compliance issue. Exceeding the regulatory radiological performance criteria would result in revised OBA radiological analyses, and could lead to license amendment requests to lower the MSIV allowed leakage value in technical specifications (resulting in more frequent rework of the valves which incurs added actual dose to workers), or may potentially require other design changes to the facility. Furthermore, these changes would be required to address delayed actuation of ECCS, an occurrence that may represent s beyond design-basis condition.

In 2009, the NRC issued for public comment a draft RG, Alternative Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, temporarily identified by its task number, DG-1199, which was the proposed Revision 1 of RG 1.183.10 DG-1199 included the proposed SAND2008-6601 MSIV leakage model which received one staff non-concurrence, a number of 9

ADAMS Accession Number 1° Federal Register Notice, 74 FR 52822

opposing internal staff comments, and a number of public comments. , When 11 12 assessing the initial dose consequence from MSIV leakage, DG-1199 did not allow credit for safety-related systems to distribute the deterministic fuel melt source term into the containment atmosphere. Rather DG-1199 assumed that the initial source of MSIV leakage is the reactor vessel (which would have a higher radionuclide concentration during the initial phase due to the assumed lack of mixing). After the end of the early in-vessel release phase, DG-1199 assumed that the source term is mixed in the containment atmosphere. The non-concurrence was concerned with the technical basis of the model and the potential for staff to impose forward fitting actions regarding certain safety-related leakage control systems for the MSIVs which had generally been approved to be taken out of service. Assuming that the reactor vessel steam dome is the initial source of MSIV leakage was felt to result in an unnecessary regulatory burden without adding a significant increase in safety. The public comments; often cited or inferred certain regulations that the model guidance would not be consistent with, had concerns with disregarding precedents on performing OBA radiological consequence analyses, raised potential regulatory issues applying the SAND2008-6601 model which would require licensees to not credit safety-related ESFs SSCs with specifically identified functional goals, and described potential impact to the facilities' probabilistic risk assessment models.13 Following the closure of the public comment period, NEI submitted a late comment 14 with an analysis performed by Metcalf (Metcalf, 2010) which used the MAAP4 code, the industry equivalent to NRC's MELCOR code, to perform similar analyses. The Metcalf paper was intended to supplement and clarify the insights presented in DG-1199 and SAND2008-6601. Metcalf s analysis utilized the industry severe accident computer code MAAP4 because of its ability to model counter-current, density-driven flow, since it was thought to be important when considering low flow leakage through the long lengths of main steam line piping. At that time, the MELCOR code did not include such modeling capability. While Metcalf s analysis was found to have some weaknesses, 15 results demonstrated that the radioactivity concentration at a point immediately upstream of an intact BWR MSIV is substantially lower than that in the reactor vessel steam dome during the time that the steam dome concentration exceeds that of the drywell. Results contradicted the conclusion reached in DG-1199 regarding the need for dose rate multipliers to compensate for the activity concentration in the steam dome being greater than that in the drywell. The Panel learned that at this point in time, the staff did not have the resources to formally address the Metcalf paper and did not pursue reconciling the two analyses.

11 ADAMS Accession Number ML091520056 for non-concurrence and an 13 The Sandia MELCOR report now re_i:iulres plants to analyze the LOCA with no ECCS injection for the first two hours. This will change the PRA analysis, impacting the large ear ly release. Is a re-evaluation of the PRA analysis now required based on this new assumption?

14 ADAMS Accession Number ML102380174 "Letter from Ralph L. Andersen (NEI) to NRG: "Nuclear Energy Institute Comments on U.S. Nuclear Regulatory Commission Draft Regulatory Guide DG-1 199, "Alternative Radiological Sources for Evaluating Design Basis Accidents at Nuclear Power Reactors,"

(Federal Register of October 14, 2009, 74 FR 52822) dated January 20, 2010.

15 ADAMS Accession Number ML19094A305

Management of Backfitting, Forward Fitting, Issue Finality, and Information Requests, Technical Specifications, source

product release from the core. Given the defense-in-depth afforded by an assumption of a significant core damage source term, it is reasonable to question if overlaying additional conservatisms to maximize offsite radiological dose, such as delaying operation of ECCS, depart from a more risk-informed perspective.

Although the NUREG-1465 source term was built upon insights gained from probabilistic risk assessment studies such as NUREG-1150, significant advancements in risk management and safety have continued to occur since 1 O CFR 50.67 was promulgated in 1999. These advancements have included:

• The increased use of probabilistic risk assessment to manage on-line risk consistent with requirements of 10 CFR 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants";

• Incorporation of additional beyond design-basis accident mitigation requirements such as those required by 10 CFR 50.54.hh for potential aircraft threats, 10 CFR 50.150 for aircraft impact assessment, and 10 CFR 50.155 for mitigation of beyond design-basis events;

• Implementation of a more risk-informed reactor oversight process for the operating reactor fleet and significant enhancements to the agency risk assessment capabilities including the SAPHIRE code and Standardized Plant Analysis Risk Models; and,

• Additional assessment of severe-accident mitigation alternatives in support of license renewal evaluations.

In addition, since the TID-14844 source term was first used to support licensing of the current fleet of nuclear power plants, a number of initiatives have reduced the need to provide additional margin for design-basis offsite radiological consequence analysis. These include regulatory requirements for beyond design-basis events (e.g., station blackout, anticipated transient without scram) and systematic searches for vulnerabilities per GL 88-20 (IPE/IPEEEE). More recent activities such as SOARCA and the ongoing Level 3 PRA project have served to further the agency's understanding of severe-accident behavior and likelihood. This context has provided an opportunity to re-evaluate if conservatisms applied in the past to address areas of high uncertainty are still required to provide adequate protection of public health and safety.

This should also be counterbalanced with the costs (including radiation exposure) that comes with maintaining plant SSCs to meet conservative analytical assumptions.

3.2 Disposition of Concerns from the Summary of Issues 3.2.1 Meeting intent of regulation at hand The regulation requires an analysis of a large release to an intact containment and does not specify an accident scenario, but RG 1.183 describes analysis assumptions for a LOCA which leads to: (1) an inappropriate need to impose non-physical assumptions pertaining to the ope.ration of safety-related SSCs to mechanistically account for the OBA source term; (2) inconsistency with LOCA analysis for 10 CFR 50.46; and (3) over-conservative assumptions intended to maximize fission product release (e.g., RG 1.183 assumptions on Mark 111 drywell to containment mixing).

source term issue transport issue source term transport source term transported transported

design-basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier."

As such, the regulations support the analysis assumption that safety-related ESF SSCs needed to perfom, a OBA prevention or mitigation function are available to perform that safety-related function. This would also support the crediting of safety-related ESF SSCs in the modeling of the transport of the source term through the nuclear power plant. Furthennore, the crediting of safety-related SSCs is fully consistent with RG 1.183, Regulatory Position 5.1.2, which allows credit for these SSCs provided certain requirements are met. However, the imposition of other requirements pertaining to the distribution and mixing of the source term (e.g., RG 1.183, Appendix A, position 3. 7 pertaining to mixing within the containment) can lead to contradictory treatment of safety-related SSC operation. Additionally, the regulatory requirements in 10 CFR 50.34, 10 CFR 50.67, 10 CFR 100.11, and the various sections of 10 CFR 52 do not require that safety-related SSCs be modeled in a manner that departs from standard design-basis assumptions. On the contrary, withholding credit for these SSCs would drive the analysis into the "beyond design-basis" regime, at which point it becomes unclear what limits the staff from arbitrarily withholding credit for other safety-related SSCs in the analysis. With regard to 10 CFR 50.46, "Acceptance criteria for ECCSs for light-water nuclear power reactors," the objective of the analysis to ensure that safety related SSCs provide sufficient mitigative functionality to preclude severe core damage (as determined by meeting the 50.46(b) acceptance criteria). Therefore, withholding credit for the functionality of safety-related SSCs to mechanistically justify a core damage source term is contrary to the objectives of 10 CFR 50.46.

Therefore, the Panel agrees with the DPO that the application of certain provisions in RG 1.183 can lead to an inappropriate application of assumptions pertaining to the operation of safety-related SSCs to mechanistically account for the source term that depart from design-basis requirements; (2) are inconsistent with LOCA analysis for 10 CFR 50.46; and (3) can be over-conservative, and in some cases non-realistic, for the purposes of "maximizing the postulated radiological consequences" as directed in RG 1.183.

It is the Panel's belief that the Imposition of a core damage source term provides sufficient defense-in-depth to assess, as noted in 10 CFR 50.34 for example, the safety features "engineered into the facility and those barriers that must be breached as a result of an accident before a release of radioactive material to the environment can occur." To defeat these same safety features for the purposes of maximizing offsite radiological dose appears contradictory to regulatory objective of assessing the capability of the plant design to mitigate an accident. Further, the imposition of assumptions that defeat safety features that can otherwise be credited can result in an arbitrary approach that lacks technical defensibility, particularly as our knowledge of severe-accident behavior continues to improve.

3.2.2 Process issue Ongoing licensing issues resulting from the staff proffering physical explanations for the deterministic source term. For example, licensees. have been instructed, via the RAI process, to delay the mixing of the drywell and suppression pool free air volumes in any type of BWR (not just those with Mark Ill containments, as noted in RG 1.183, Appendix A,

source term source term

assumption is based on the time period of approximately two hours, as determined in NUREG-1465, for the in-vessel source term to fully develop. The argument for the use of this assumption is that the substantial fuel melt source term could not develop if safety-related systems operated as designed. However, this argument contradicts the deterministic use of the substantial fuel melt source term and credit for safety-related SSCs in design-basis analyses.

Although the Panel did identify some examples where the RAI process was used to imply that these assumptions should be used, the Panel did not identify widespread use of RAI to impost this modeling assumption. It appears that the first instance was during the Perry review when the staff challenged the licensee's initial assumption of mixing at T = 0.20 Perry changed their analysis to credit mixing after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. After that it appears that licensees followed the Perry precedent with the exception of the Susquehanna license amendment request (LAR) which generated a RAI and a subsequent re analysis.21 The Panel did not find any other licensee that attempted to deviate from the Perry precedent. However, the DPO Panel did find, as evidenced in DG-1199 Regulatory Position A-5, the staff were intending to provide physical explanations for the deterministic source term that match the AST technical basis by referencing for use SAND2OO8-66O1 as an acceptable model for estimating the radioactivity available for release via MSIV leakage. This method uses the NUREG-1465 source term in conjunction with a recommended source term fraction to be placed in the steam dome volume during the first two hours of the accident. As explained in SAND 2008-6601, [i]n order for the model to represent the physical flow paths of the main steamline connections, the source volume is given the volume of the steam dome for the first two hours (after 2 hrs it is assumed that the vessel has been reflooded with an assumed equilibration of steam dome and drywell fission product concentrations)." To model this assumption, the licensee would need to assume the safety-related equipment would not function for some predete.rmined period.

The DPO Panel disagrees the staff instructed licensees through the standard regulatory review process to unwillingly adopt these modeling practices. While the precedent established under the Perry AST review may have been, in some areas, over1y conservative, the DPO Panel feels both staff and licensees utilized precedence appropriately.

3.2.3 Crediting safety-related SSCs in analyses Inconsistent treatment of crediting safety-related SSCs within the deterministic radiological consequence analyses when determining compliance with the various OBA radiological performance criteria.

Response: Design basis accident analyses performed for purposes other than radiological consequence assessment credit operation of safety-related SSCs to prevent 20 See Perry Safety Evaluation Report, dated March 26 1999 (ADAMS Accession Number ML021840462), and Supplemental Information, dated January 18, 1999 (ADAMS Accession Number ML20199G615) 21 See SusQuehanna Safety Evaluatron Report dated January 31, 2007 (ADAMS Accession Number ML070080301), and the original license amenament request to adopt 50.67 (ADAMS Accessfon Number ML060120353)

or mitigate the event. For example, the definition of safety-related SSCs in 10 CFR 50.2 specifically notes that these SSCs are "relied upon to remain functional during and following design-basis events" in order to assure, in part, the "capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures." The radiological consequence assessment to show compliance with regulatory radiological performance criteria, as described in the regulations identified above, necessarily assumes failure of the fuel, cladding and RCS as fission product barriers with the purpose of evaluating the fission product release mitigating design features (e.g. containment). Imposing explanations of the physical reasons for the core damage that drives source term assumptions could result in inconsistent crediting of SSCs within the radiological consequence analyses, and also in comparison to other deterministic safety analyses for the facility. Given the failure of the initial fission product barriers associated with the fuel system must be assumed to be consistent with the*

regulatory footnote on fission product release, the reason for such failures can be considered to be immaterial to the evaluation of the radionuclide release barriers that provide additional layers of defense-in-depth. The operation and capability of ESF systems and safety-related design features that provide radionuclide release mitigation, as modeled for the purposes of radionuclide transport, retention and removal within the containment and other release pathways to the environment, should be credited consistent with other design-basis accident safety analyses. In this sense, the Panel would support the DPO proposed resolution of this issue to "over1ay" the source term onto the designed operation of the facility during the design-basis accident.

In addition, considering*the issues the DPO Submitter raised, the Panel supports providing additional clarification in the guidance for radiological consequence analysis to ensure that provisions related to crediting safety-related SSCs appropriately consider their assurance of safety functions and the design bases of the facility.

4. CONCLUSIONS Current regulatory practices for design-basis offsite consequence analysis have included certain mechanistic system performance assumptions (consistent with a core damage source term) in order to maximize offsite radiological consequences. Although these assumptions are not specifically required by regulations, they have provided additional margin to account for severe accident uncertainties and were intended to ensure the protection of public health and safety.

However, these assumptions have led the staff to focus, in some cases, on low likelihood severe-accident scenarios (e.g., large-break LOCA coincident with ear1y failure followed by later recovery of ECCS) and apply system response assumptions that are inconsistent with other design and licensing basis assumptions. This has led to confusion over licensing basis requirements and ambiguity in the assumptions needed to support these analyses.

The DPO Panel concluded that current regulatory requirements do not require an applicant or licensee to apply mechanistic system performance assumptions for design-basis consequence analysis to justify the use of a core damage source term for design-basis offsite consequence analysis. Instead, the source term should be applied assuming nominal design-basis safety system capability (including design-basis assumptions for single-failure, availability of power, etc.). This position is consistent with RG 1.183, Regulatory Position 5.1.2, "Credit for Engineered Safeguard Features," but may be inconsistent with other provisions in RG 1.183 related to source term mixing and distribution guidance.

5. RECOMMENDATIONS

1. The guidance for the design-basis accident radiological consequence analysis should be revised to clarify that safety-related SSCs meeting certain criteria can, and should, be credited. Guidance should also be reviewed to ensure that regulatory positions do not inadvertently result in disallowing credit for safety-related SSCs when it would be otherwise appropriate. As part of the revision of RG 1.183, MHA and LOCA terminology should be clarified to ensure consistency with meeting the regulatory objective.

By removing the motivation to attempt to mechanistically explain the specified source term by defeating certain safety-related functions, the evaluation of the radiological consequences from an MHA would be simplified. The source term for the MHA is, and always has been, deterministic or non-mechanistic. Regulations do not require an explanation for how substantial fuel melt occurs, only that it is based on a major accident that is generally been assumed to result in substantial meltdown of the core. By not altering normal design-basis assumptions to provide an explanation for how the source term is derived, this approach ensures that SSC functionality remains consistent with the design and licensing basis. Therefore, staff should model the MHA with the deterministic substantial fuel melt source term being injected or overlaid into the radiological consequence analysis notwithstanding the operation of safety-related equipment designed to preclude significant fuel failure. The purpose of this approach would be to test the adequacy of the containment and other safety-related systems. All other aspects of a plant's response to the accident should be credited if the systems are safety-related, as this designation ensures reliability to perform their safety function.

Safety-related pumps and valves need to be credited as performing their design functions including those systems designed to deliver water to flood the core, thereby mixing the deterministic fuel melt source term into the containment atmosphere. The additional measure of defense-in-depth is gained by non-mechanistically assuming that despite the operation of these systems substantial fuel melt occurs.

2. The Panel recommends providing clarification to the staff when developing guidance for new radionuclide transport models to reinforce the Commission direction from SRM SECY-96-242, Use of the NUREG-1465 Source Term at Operating Reactors, where the Commission advised:

"The staff should exercise caution so as to avoid creating new severe-accident mitigation requirements in the licensing of currently operating plants that do not have explicit and informed Commission approval."

This would assist the staff when developing new and improved models. Acknowledging the radiological consequence analyses are intended to maximize consequences within the design basis while preserving a reasonable approach to safety.22

3. The Panel recommends the staff to consider clarifying the Standard Review Plan (NUREG-0800) and applicable Regulatory Guidance with regard to crediting safety related SSCs in the radiological consequence analyses.

22 See memo from Ho Nieh to All Staff who Support the Nuclear Reactor Safety Program, "Applying the Principles of Good Regulation as a Risk-informed Regulator," (ADAMS Accession No. ML19260E683)

4. Guidance related to the use of precedent licensing reviews should be enhanced to note that the staff should take care in applying precedence and ensure that the use of past regulatory and licensing precedents remain valid given the current knowledge state and do not impose unnecessarily conservative assumptions or criteria on licensees.

5. The DPO Panel believes that RG 1.183, Revision 0, is time-tested and provides an acceptable, albeit conservative, method to address design-basis accident radiological consequences. Therefore, the Panel believes that the RG can continue to be used by licensees. However, the staff should avoid imposing the guidance contained in RG 1.183, Revision 0, if it would result in unnecessary conservatism that is inconsistent with regulatory requirements.

6. The Office of Nuclear Regulatory Research should be consulted to assess (1) if performance of an integrated analysis, rather than a piece meal approach of assessing each radiological release pathway individually, would result in a more realistic and feasible approach that avoids inconsistent analytical assumptions, and (2) if code enhancements are needed for severe accidents codes to permit overlaying of the core damage source term without the need to defeat safety-related SSCs23.

23 The Panel determined that part of the motivation for disabling certain safety-related features in thermal hydraulic modeling (e.g., in MELCOR) was to allow the code's severe accident modeling mechanistically develop a source term equivalent to the source term specific in NU REG-1465. If this equipment were to continue to operate, the thermal hydraulic codes would not develop significant core damage. Therefore, it may be necessary to develop enhanced code capability to allow the severe accident source term to be overlaid with expected design basis event SSC functfonallty.

Analysis of Main Steam Isolation Valve Leakage in Design Basis Accidents Using MELCOR 1.8.6 and RADTRAD (SAND2008-6601) (ADAMS Accession No. ML083180196).

BWR Steam Line Radionuclide Concentration Distribution following a DBA LOCA.

Nuclear Reactor Safety: On the History of the Regulatory Process.

Accident Source Terms for Light-Water Nuclear Power Plants Using High-Burnup or MOX Fuel (SAND2011-0128).

Calculation of Distance Factors for Power and Test Reactors (TID-14844) (ADAMS Accession No. ML021720780).

Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants (NUREG-1150).

Issues Pertaining to Evolutionary and Passive Light-Water-Reactor Designs" (ADAMS Accission No. ML003708141.

Accident Source Terms for Light-Water Nuclear Power Plants (NUREG-1465).

SRM SECY-96-242: Use of the NUREG-1465 Source Term at Operating Reactors (ADAMS Accession No. ML003752965).

SECY-98-154: Results of the Revised NUREG-1465 Source Term Rebaselining for Operating Reactors (ADAMS Accession No. ML15309A319).

Federal Register: Use of Alternative Source Terms at Operating Reactors (64FedReg71990).

Regulatory Guide 1.183: Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors (ADAMS Accession No. ML003716792).

NRC Regulatory Issue Summary 2006-04, Experience with Implementation of Alternate Source Terms March, 7, 2006 (RIS 2006-04).

Draft Regulatory Guide DG-1199: Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors (ADAMS Accession No. ML090960464).

State-of-the-Art Reactor Consequence Analyses (SOARCA) Report (NUREG-1935).

SECY-16-0012: Accident Source Terms and Siting for Small Modular Reactors and Non-light Water Reactors (ADAMS Accession No. ML15309A319).

SRM SECY-19-0036: Application of the Single Failure Criterion to NuScale Power LLCs Inadvertent Actuation Block Valve (ADAMS Accession No. ML19183A408).

Memo From Michael Case to Joseph Donoghue and Michael Franovich, "Applicability of Source Term for Accident Tolerant Fuel, High Burn up and Extended Enrichment" (ADAMS Accession No. ML20126G376).

Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition (NUREG-0800, Formerly issued as NUREG-75/087).

Appendix A: Assessment of staff safety evaluation reports approving 10 CFR 50.67 AST license amendment requests This appendix provides a high-level assessment of staff safety evaluation reports approving 10 CFR 50.67 AST license amendment request regarding the assumption of drywell/wetwell mixing for boiling water reactors after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The discussions below are generally direct excerpts from the reports with some commentary. It appears the first instance of the assumption of drywell/wetwell mixing after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> was during the initial Perry pilot review when staff challenged the licensee's initial assumption of mixing at T :: 0 which has a Mark Ill containment. Perry changed their analysis to credit mixing after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Subsequent license amendment requests appear to follow the Perry precedent, regardless of their containment design, except for the Susquehanna license amendment request which generated a set of RAls resulting in a subsequent re-analysis by the licensee. The subsequent re-analysis resulted in little difference in computed radiological consequences. No licensees' appeared to attempt to deviate from the Perry precedent presumably because of regulatory certainty. Based on the wording in the safety evaluation reports, it appears that (with the exceptions of Perry and Susquehanna) the licensees made the 2-hour delay assumption in their initial license amendment requests.

Perry SER March 26, 1999 (ADAMS Accession No. ML021840462)

From the attached Supplemental Information (ADAMS Accession No. ML20199G615):

Changes To The revised accident source terms Calculations The calculations included in Attachment 3 were revised to reflect resolution of several items with the NRC staff. Note that these issues are specific to the Perry Nuclear Power Plant (PNPP).

Several of these items are expected to receive additional attention as part of the generic RG development in support of the current rulemaking effort. It is possible that if PNPP revises revised accident source terms calculations in support of future applications, the PNPP values may change in such future calculations as a result of the generic industry discussions. The primary changes are summarized below. Further details on these changes are described in the attached calculations.

• The sweepout flow rate from drywell to containment during the fission product release phase of the LOCA event (approximately the first two hours) is assumed to be 3000 cubic feet per minute (cfm) in the thermal-hydraulic calculation. This was reduced by more than half from the previous calculation, from ~6200 cfm. Although there are a number of mechanisms that would produce sweepout flow rates greater than or equal to the 6200 cfm sweepout rate, this change was made in order to increase the calculated dose from the main steam lines during the first two hours of the event.

• After two hours, the drywell and the unsprayed region of the containment are assumed to be well-mixed. The ECCS injection function is assumed to be recovered at the two hour point, which is an integral assumption for application of the revised accident source terms. The previous calculations assumed that when the Emergency Core Cooling Systems began injecting, the resultant steam generation spike would sweep essentially all of the source term out of the drywell and into the unsprayed region of the wetwell (the containment). [emphasis added] This change to a well-mixed volume was made so that

the mitigation techniques for the leakage from both the containment and the drywell would be more equally challenged in the time frame after the ECCS injection recovery at two hours.

Grand Gulf SER March 14, 2001 (ADAMS Accession No. ML010780172)

As characterized in NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants," the gap and early in-vessel fission product releases terminate two hours after the postulated LOCA initiation. The staff assumed (as it did for Perry Nuclear Power Plant, Unit 1 (Perry), License Amendment No.103, issued on March 26, 1999, "Main Steam Line Leakage Requirements and Elimination of the Main Steam Isolation Valve Leakage Control System Implementing the Alternative Source Term") that the fission products are homogeneously distributed between the drywell and the primary containment two hours after accident initiation (both GGNS and Perry containment structures are Mark Ill type designs). This would require reflooding of the reactor vessel. Instead of trying to justify an all encompassing steaming rate due to this reflooding, the staff concludes that a substantial amount of fission products may end up in the primary containment as well as the drywell, and that mitigative features such as the SGTS need to be designed to accommodate a significant portion of the source term. For most of the risk significant cases, such as station blackout and transients, all the fission products are released directly to the primary containment via the safety relief valves. Waiting two hours after accident initiation to homogeneously mix the source term is acceptable for achieving an appropriate balance because the worst two hours are considered, instead of the first two hours used with the TID-14844 source term.

Confirmatory calculations performed by the staff showed that the radiological consequences are dependent upon the drywell bypass leakage prior to the termination of fission product release at two hours. Because of this sensitivity, the staff concludes, as it did for Perry, that without relocation of reactor fuel to the lower head, the steaming rate of an intact core on the order of 3,000 cfm should be assumed for drywell bypass leakage. The staffs steaming rate prior to two hours is conservative in that it does not credit steaming due to relocation, cooling from alternative water sources (AWS}, or the release of hydrogen gas.

Duane Arnold SER July 31, 2001 (ADAMS Accession No. ML011660142)

The LOCA considered in this evaluation is a complete and instantaneous severance of one of the recirculation loops. The pipe break results in a blowdown of the reactor pressure vessel (RPV) liquid and steam to the drywell via the severed recirculation pipe. The resulting pressure buildup drives the mixture of steam, water, and other gases down through vents to the downcomers and into the suppression pool water thereby condensing the steam and reducing the pressure. Due to the postulated loss of core cooling, the fuel heats up, resulting in the release of fission products. Under the TID14844 (J.J. DiNunno, et al., Calculation of Distance Factors for Power and Test Reactor Sites, USAEC TID-14844, U.S. Atomic Energy Commission, 1962) assumption of instantaneous core damage, this initial blowdown would also include fission products, a fraction of which would be retained by the suppression pool water.

Under the AST, the fission product release occurs in phases over a two-hour period. Significant quantities of fission products would not be part of the initial blowdown to the suppression pool.

Subsequent recirculation of suppression pool water by the ECCS would cause some transport

of fission products between the drywell and the wetwell, and some scrubbing effect. NMC has conservatively assumed no credit for suppression pool scrubbing of fission products.

In addition, NMC has conservatively assumed that the fission product release from the RPV is homogeneously dispersed within the drywell free volume only, ignoring the free volume of the wetwell. [emphasis added]

Brunswick SER May 30, 2002 (ADAMS Accession No. ML021480483)

The LOCA considered in this evaluation is a complete and instantaneous severance of one of the recirculation loops. The pipe break results in a blowdown of the RPV liquid and steam to the drywall via the severed recirculation pipe. The resulting pressure buildup drives the mixture of steam, water, and other gases down through vents to the downcomers and into the suppression pool water thereby condensing the steam and reducing the pressure. Due to the postulated loss of core cooling, the fuel heats up, resulting in the release of fission products. Under the traditional TID-14844 assumption of instantaneous core damage, this initial blowdown would also include fission products, a fraction of which would be retained by the suppression pool water. Under the AST, the fission product release occurs in phases over a 2-hour period.

Significant quantities of fission products would not be part of the initial blowdown to the suppression pool. Subsequent recirculation of suppression pool water by the ECCS would cause some transport of fission products between the drywell and the wetwell, and some scrubbing effect. CP&L has conservatively assumed no credit for suppression pool scrubbing of fission products. In addition, CP&L has conservatively assumed that the fission product release from the RPV is homogeneously dispersed within the drywall free volume only, ignoring the free volume of the wetwell. [emphasis added]

River Bend SER March 14. 2003 (ADAMS Accession No. ML030760746)

As characterized in NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants, the gap and early in-vessel fission product releases terminate 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the postulated LOCA initiation. The NRC staff assumed [as it did for Perry Nuclear Station (Perry)]

License Amendment No.103, issued on March 26, 1999, Main Steam Line Leakage Requirements and Elimination of the Main Steam Isolation Valve Leakage Control System Implementing the Alternative Source Term, and for Grand Gulf License Amendment No.145, issued on March 14, 2001, Fu/I-Scope Implementation of an Alternative Accident Source Term,) that the fission products are homogeneously distributed between the drywell and the primary containment 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after accident initiation. This would require reflooding of the reactor vessel. Instead of trying to justify an all encompassing steaming rate due to this reflooding, the staff concludes that a substantial amount of fission products may end up in the primary containment as well as the drywell. For most of the risk significant cases, such as station blackout and transients, all the fission products are released directly to the primary containment via the safety relief valves. Waiting 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to homogeneously mix the source term is acceptable for achieving an appropriate balance because the worst 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> are considered, not the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> used with the TID-14844 source term.

Confirmatory calculations performed by the NRC staff showed that the radiological consequences are dependent upon the drywall bypass leakage prior to the termination of fission product release at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Because of this sensitivity, the staff concludes, as it did for

Perry and Grand Gulf, that, without relocation to the lower head, the steaming rate of an intact core on the order of 3,000 cubic feet per minute (cfm) should be assumed for drywell bypass leakage. The staff's steaming rate prior to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is conservative in that it does not credit steaming due to relocation, cooling from AWSs, or the release of hydrogen gas.

The 3,000 cfm drywell bypass leakage rate is based upon large-break LOCA analyses performed with MELCOR on a Grand Gulf type model. These analyses showed no relocation below the core plate, water level below the core plate, and an average steaming rate of approximately 2,800 cfm prior to quenching of the core at approximately 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. Also, AWSs, such as the standby liquid control system, would not be available during station blackout sequences, which comprised 96 percent of the core damage frequency in the Grand Gulf case.

Therefore, the NRG staff concludes the use of 3.000 cfm for the drywell bypass leakage prior to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is also reasonable for RBS. [emphasis added}

Hope Creek SER April 15, 2003 (ADAMS Accession No. ML030760293); LAR June 28, 2002 (ADAMS Accession No. ML021980018); RAI Response December 18, 2002 (ADAMS Accession No. ML023610579); Re-Submittal based on RAI January 18, 2003 (ADAMS Accession No. ML030290614); RAI response February 25, 2003 (ADAMS Accession No. ML030630623) Docket# 05000354 The safety evaluation report is silent on drywell/wetwell mixing; RAls silent on drywell/wetwell mixing Browns Ferry SER September 27, 2004 (ADAMS Accession No. ML042730028)

The LOCA considered in this evaluation is a complete and instantaneous severance of one of the recirculation loops. The pipe break results in a blowdown of the RPV (RPV) liquid and steam to the drywell via the severed recirculation pipe. The resulting pressure buildup drives the mixture of steam, water, and other gases down through vents to the downcomers and into the suppression pool water thereby condensing the steam and reducing the pressure. Due to the postulated loss of core cooling, the fuel heats up, resulting in the release of fission products.

Under the TID-14844 assumption of instantaneous core damage, this initial blowdown would also include fission products, a fraction of which would be retained by the suppression pool water. Under the AST, the fission product release occurs in phases over a 2-hour period. TVA has conservatively assumed that the fission product release from the RPV is homogeneously dispersed within the drywell free volume only for the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. [emphasis added] TVA assumes that core quenching occurs at about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> resulting in substantial steam production in the RPV and drywell that will purge a large fraction of the drywell atmosphere through the torus downcomer vents, through the suppression pool, and into the torus airspace. TVA did not credit any reduction in fission products transferred to the torus air space by suppression pool scrubbing. Instead, TVA assumes a well-mixed torus air space and drywell.

Fermi 2 SER September 28, 2004 (ADAMS Accession No. ML042430179)

The LOCA considered in this evaluation is a complete and instantaneous severance of one of

the two recirculation loops resulting in a blowdown of the RPV liquid and steam to the drywell.

Because the previous TIO 14844, dated March 23, 1962, specified a source term assumption of instantaneous core damage, this initial blowdown would also include all of the released fission products, a fraction of which would be retained by the suppression pool water. Under the AST, a substantial fraction of the fission product release occurs after the initial blowdown is complete.

As such, the licensee did not credit any reduction in fission products transferred to the torus air space by suppression pool scrubbing, assuming instead a well-mixed torus air space and drywell. The NRC staff finds the licensee's assumptions in this area acceptable.

The licensee's revised analysis assumes that the released fission products are dispersed throughout the drywell free volume, that there is no mixing of the drywell and wetwell volumes for the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and that there is complete mixing after that. The NRC staff finds this assumption acceptable as the AST is effectively based on a terminated LOCA in which core cooling is restored at the end of the early in-vessel release phase. The licensee assumes an MSIV leak rate of 100 scfh in steam line B, 50 scfh in steam line D, and no leakage in remaining steam lines A and C for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 50 percent of these values for the next 29 days. Steam lines B and D are each modeled as two deposition nodes - the RPV nozzle to the inboard MSIV, and the inboard MSIV to the third MSIV. Since a main steam rupture within the primary containment could be an initiator for a LOCA, the licensee assumes no deposition credit for the inboard node in steam line B. In keeping with single-failure considerations, the outboard MSIV on steam line B is assumed to fail open.

Clinton SER September 19, 2005 {ADAMS Accession No. ML052570461)

The LOCA considered in this evaluation is a complete and instantaneous severance of one of the recirculation loops. The pipe break results in a blowdown of the reactor pressure vessel (RPV) liquid and steam to the drywell via the severed recirculation pipe. The resulting pressure buildup drives the mixture of steam, water, and other gases through the suppression pool water and into the primary containment. The suppression pool water condenses the steam and reduces the pressure. After the initial RPV blowdown, ECCS water injected into the RPV will spill into the drywell, transporting fission products to the suppression pool and then into the primary containment. In lieu of modeling this transport mechanistically, AmerGen has conservatively assumed that the fission product release from the fuel is homogeneously and instantaneously dispersed within the drywell free volume, with 3000 cubic feet per minute (cfm) of the drywell air flowing to the primary containment for the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Rapid mixing (108 cfm) between the drywell and primary containment is assumed after the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for the duration -

of the 30-day accident to model that the fission products are homogeneously distributed between the drywell and primary containment. The staff finds that the licensee's assumptions regarding drywell and containment mixing are consistent with assumptions previously found acceptable for a full implementation of an AST for the Perry Nuclear Power Plant. which has a Mark Ill containment such as CPS. [emphasis added] AmerGen did not credit any reduction by suppression pool scrubbing for fission products transferred to the primary containment through the suppression pool.

Limerick SER August 231 2006 (ADAMS Accession No. ML062210214)

The objective of analyzing the radiological consequences of a LOCA is to evaluate the design of various plant safety systems. These safety systems are intended to mitigate the postulated release of radioactive materials from the plant to the environment in the event that the ECCS is not effective in preventing core damage. A LOCA is a failure of the reactor coolant system (RCS) that results in the loss of reactor coolant which, if not mitigated, could result in fuel damage, including a core melt. The primary coolant blows down through the break, depressurizing the RCS. As the pressure builds in the drywell, steam and other gases expand into the wetwell. While passing through the suppression pool water, the steam is condensed, thereby reducing the pressure in the wetwelt and drywell. A reactor trip occurs and the ECCS actuates to remove fuel decay heat. Thermodynamic analyses, performed using a spectrum of RCS break sizes, show that the ECCS and other plant safety features are effective in preventing significant fuel damage. Nonetheless, the radiological consequence portion of the LOCA analysis assumes that the ECCS is not effective and that substantial fuel damage occurs.

Appendix A of RG 1.183 identifies acceptable radiological analysis assumptions for a LOCA.

The source term and release pathways related to the LOCA are discussed below ..... .

Under the previously used TIO 14844 source term assumption of instantaneous core damage and fission product release, the initial blowdown would also include all of the released fission products, a fraction of which would be retained by the suppression pool water. Under the AST, a substantial fraction of the fission product release from the core occurs after the initial blowdown is complete. Therefore, Exelon did not credit any reduction in fission products transferred to the wetwell air space by suppression pool scrubbing, assuming instead, a well mixed wetwell air space and drywell after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Dresden/Quad SER September 11, 2006 (ADAMS Accession No. ML062070290}

Exelon assumed that the fission products released from the RCS following the postulated LOCA are instantaneously and homogeneously mixed throughout the free air volume of the drywell for the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. As characterized in NUREG-1465, the fission product releases (gap and early-in-vessel releases) terminate 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the onset of the postulated LOCA.

This would require retlooding of the RPV. Exelon asserts, and the NRC staff agrees, that reflooding and core quenching would occur at about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> resulting in substantial steam production in the RPV that will purge a large fraction of the drywell atmosphere through the drywell main vent, the main vent header, eight torus vent pipes, 96 downcomers arranged in pairs and submerged in the suppression pool water, the suppression pool water, wetwell air space, and finally back to the drywell through vacuum breaker lines.

The NRC staff believes that the mass and energy (steaming and steam condensation) created by reflooding (arresting RPV failure) and core quenching will provide sufficient energy to mix the drywell and wetwell air when vacuum breaker cycling occurs during this pressure transient. The licensee assumed that the radioactivity release is diluted into the larger volume of the wetwell plus drywell air spaces after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Before this time, the radioactivity is only assumed to be released into the drywell net free volume. The NRC staff expects that most fission products (other than noble gases and iodine in organic form) in the drywell air transferred to the torus air space after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> will be scrubbed by the suppression pool water. However, Exelon did not credit any reduction in fission products transferred to the torus air space from drywell by

suppression pool scrubbing. Instead, Exelon assumed a well-mixed torus air space and drywell after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Columbia SER November 26, 2006 (ADAMS Accession No. ML062610440)

Note: Accident description states that drywell/wetwell mixing occurs with accident initiation but 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> delay is assumed to align with source term assumptions.

The objective of analyzing the radiological consequences of a LOCA is to evaluate the design of various plant safety systems. These safety systems are intended to mitigate the postulated release of radioactive materials from the plant to the environment in the event that the ECCS is not effective in preventing core damage. A LOCA is a failure of the reactor coolant system (RCS) that results in the loss of reactor coolant that, if not mitigated, could result in fuel damage, including a core melt. The primary coolant blows down through the break to the drywell, depressurizing the RCS. As the pressure builds in the drywell, steam and other gases expand into the wetwell. Passing through the suppression pool water, the steam is condensed, thereby reducing the pressure in the wetwell and drywell. [emphasis added] A reactor trip occurs and the ECCS actuates to remove fuel decay heat. Thermodynamic analyses, performed using a spectrum of RCS break sizes, show that the ECCS and other plant safety features are effective in preventing significant fuel damage. Nonetheless, the radiological consequence portion of the LOCA analysis assumes that ECCS is not effective and that substantial fuel damage occurs. Appendix A of RG 1.183 identifies acceptable radiological analysis assumptions for a LOCA. The source term and release pathways related to the LOCA are discussed below.

Under the previous TID-14844 source term assumption of instantaneous core damage, the initial blowdown would also include all of the released fission products, a fraction of which would be retained by the suppression pool water. Under the AST, a substantial fraction of the fission product release occurs after the initial blowdown is complete. As such, EN did not credit any reduction in fission products transferred to the wetwell air space by suppression pool scrubbing, assuming instead a well-mixed wetwell air space and drywell after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. [emphasis added]

Monticello SER December 7, 2006 (ADAMS Accession No. ML062790015}

The LOCA pipe break results in a blowdown of the reactor pressure vessel (RPV) liquid and steam to the drywell. The resulting pressure buildup drives the mixture of steam, water, and other gases through the suppression pool water and into the primary containment. The suppression pool water condenses the steam and reduces the pressure. After the initial RPV blowdown, ECCS water injected into the RPV will spill into the drywell, transporting fission products to the suppression pool and then into the primary containment. In lieu of modeling this transport mechanistically, the licensee conservatively assumed that the fission product release from the fuel is homogeneously and instantaneously dispersed within the drywell free volume.

Rapid mixing between the drywell and wetwell (torus) airspace is assumed after the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for the assumed duration of 30 days to model fission products being homogeneously distributed between the drywell and torus airspace. [emphasis added) The NRC staff finds that the licensee's assumptions regarding drywell and wetwell mixing are consistent with assumptions previously found acceptable for full implementation of an AST for other Mark I containments.

The licensee did not credit any reduction by suppression pool scrubbing for fission products transferred to the primary containment through the suppression pool.

Susquehanna SER January 31, 2007 (ADAMS Accession No. ML070080301) 3.1.1.1.2 Containment mixing In the submittal dated October 13, 2005 (ADAMS Accession No. ML060120353), PPL modeled the primary containment leakage pathway treating the drywell and primary containment as a single, well-mixed volume from the start of the event. This assumption may not be supportable during the early stages of the event. The initial blowdown of the reactor coolant system would have occurred prior to the onset of the in-vessel release phase. Thus, the driving force for mixing between the two volumes will be less at the time when substantial core damage is occurring. Since the LOCA break communicates with the drywell volume only, the use of the drywell and wetwell free volume has the effect of reducing the concentration of the fission products available for release from containment leakage, a non-conservative situation. Because of this uncertainty, the NRC staff has deterministically assumed that complete mixing does not occur until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> post-LOCA, when core reflood is projected as PPL assumed for the MSIV leakage pathway and the secondary containment bypass pathway. In response to a RAI, PPL updated the LOCA analysis assuming that complete mixing within the primary containment does not occur until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> post-LOCA for all pathways. The results of the updated analysis indicate that the control room (CR) and the LPZ doses are not changed from the values provided in the AST submittal. The time period for the worst case 2-hour EAB dose changed and the EAB dose from this pathway increased from 3.07 rem to 3.66 rem TEDE. PPL asserts and the NRC staff agrees that the conclusion that the EAB dose is well within the regulatory limit of 25 rem remains valid with the incorporation of the delay in complete containment mixing to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> post-LOCA.

Oyster Creek SER April 26. 2007 (ADAMS Accession No. ML071080019)

The Mark I primary containment is composed of a drywell and a wetwell/torus which includes the suppression pool. The radioactivity released from the reactor coolant system to the containment is assumed to instantaneously and homogeneously mix throughout the drywell air space. Mixing between the drywell and wetwell air space is modeled at 9180 cubic feet per minute (cfm) from 1.129 hours0.00149 days <br />0.0358 hours <br />2.132936e-4 weeks <br />4.90845e-5 months <br /> to 1.296 hours0.00343 days <br />0.0822 hours <br />4.89418e-4 weeks <br />1.12628e-4 months <br />, based on the expected steam flow from the drywell to the wetwell from analysis with the MAAP4 thermal-hydraulics code. The drywell and wetwell air space volumes are assumed well-mixed by transfer of 10 drywell volumes per hour after 2.008 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> until the end of the event, due to expected opening of the vacuum breakers between the wetwell and drywell. The NRC staff finds this formulation to be reasonably conservative and the modeling of the primary containment volume acceptable.

Nine Mile Point I SER December 19, 2007 (ADAMS Accession No. ML073230597)

For releases into containment, the licensee assumed that activity released from the reactor coolant system is well-mixed between the drywell and the torus airspace volumes, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the onset of gap release following the restoration of ECCS, which is 122 minutes after accident

initiation. Before this time, the releases are only mixed in the drywell airspace. The licensee assumed that, at the time the ECCS is restored, the thermal-hydraulic response of cooling water quenching the molten core and core debris in the PC will result in the drywell and torus airspace volumes becoming well-mixed. This assumption is acceptable for the "light bulbn and torus design of the Mark I containment at NMP1, as it is configured with downcomers from the drywell that extend below the surface of the torus suppression pool coolant (wetwell). The licensee's assumption of a one wetwell volume per minute rate of mixing is also acceptable. The licensee takes no credit for the activity decontamination, or scrubbing, associated with such activity releases into the suppression pool flufd.

Nine Mile Point II SER May 29. 2008 (ADAMS Accession No. ML081230439)

For releases into containment, the licensee assumed that activity released from the reactor coolant system is well-mixed between the drywell and the suppression chamber airspace volumes 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the onset of gap release following the restoration of ECCS, which is postulated to occur 122 minutes after accident initiation. Before this time, the releases are only mixed in the drywell airspace. The licensee assumed that, at the time the ECCS is restored, the thermohydraulic response of cooling water quenching the molten core and core debris in the PC will result in the drywell and suppression chamber airspace volumes becoming well-mixed. This assumption is acceptable for the Mark II containment design of NMP2, as it is configured with downcomers from the drywell that extend below the surface of the suppression pool coolant (wetwell). The licensee's assumption of a one wetwell volume per minute rate of mixing is also acceptable. The licensee takes no credit for the activity decontamination, or scrubbing, associated with such activity releases into the suppression chamber fluid.

Hatch SER August 28i 2008 {ADAMS Accession No. ML081770075)

For releases into containment, the licensee assumed that activity released from the reactor coolant system begins to mix between the drywell and the suppression chamber (torus) airspace volumes, at 2.03 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, approximately coincident with the hypothetically modeled restoration of ECCS, and completion of the early in-vessel activity release phase, which is postulated to be complete 122 minutes after accident initiation. Before this trme, the releases are only mixed in the drywell airspace. The licensee calculated that, at the time the ECCS is restored, the thermohydraulic response of cooling water quenching the molten core and core debris in the PC will result in the mixlng of the drywell and torus airspace volumes. This assumption is acceptable for the Mark I containment design of HNP, as it is configured with downcomers from the drywell that extend below the surface of the torus coolant (wetwell). The licensee calculates this mixing to occur at time-dependent rates after 2.03 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />; this mixing profile is shown in Table 3.2.1. The licensee took no credit for the activity decontamination, or scrubbing, associated with such activity releases into the wetwell.

Peach Bottom SER September 5. 2008 (ADAMS Accession No. ML082320406)

The radioactivity released from the fuel is assumed to mix instantaneously and homogeneously throughout the free air volume of the primary containment (drywell). The radioactivity release into the drywell is assumed to terminate at the end of the early in-vessel phase, which occurs at

2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the onset of a LOCA. The reduction in drywell leakage activity by dilution in the reactor building (RB) and removal by the SGT system filtration is not credited. The analysis dilutes the radioactivity released from the core into the drywell air volume during the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the LOCA, and then into the combined drywell plus suppression chamber air volume after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, at which time the containment volume becomes well-mixed following the restoration of core cooling. The thermal-hydraulic conditions in the primary containment are expected to be quite active at this time due to a very high flow established between the drywell and wetwell as a result of steaming and condensing phenomenon.

Cooper SER September 15. 2009 (ADAMS Accession No. ML092310349)

The licensee assumed a 2-hour delay in the startup of the ECCS after the onset of gap release, consistent with an assumption of a loss-of-offsite power (LOOP) concurrent with the design basis LOCA. This assumption was made for the purpose of attributing the onset of the deterministically defined core melt to a specific mechanism in order to remain consistent and conservative with respect to the applicable regulatory guidance and requirements.

For releases into containment, the licensee assumed that activity released from the reactor coolant system is instantaneously and homogeneously well-mixed in the drywell. This assumption is conservative and consistent with the guidance of RG 1.183. Therefore, it is acceptable to the NRC staff. In addition to complying with the regulatory guidance, the licensee takes no credit for postulated ECCS restoration and the resulting thermohydraulic response of cooling water quenching the molten core and core debris in the PC. This hypothetical phenomena is generally assumed to result in the drywell and torus airspace volumes becoming well-mixed for the "light bulb" and torus design of the Mark I containment, like that of CNS, as it is configured with downcomers from the drywell that extend below the surface of the torus suppression pool coolant (wetwell). Also, as a result of the licensee's conservative assumption, no credit is taken for the activity decontamination, or scrubbing, associated with such activity releases into the suppression pool fluid.

LaSalle SER June 61 2010 (ADAMS Accession No. ML101750625)

The LOCA considered in this evaluation is a complete and instantaneous severance of one of the recirculation loops. The pipe break results in a blowdown of the RPV liquid and steam to the drywell via the severed recirculation pipe. The resulting pressure buildup drives the mixture of steam, water, and other gases through the suppression pool water and into the primary containment. The suppression pool water condenses the steam and reduces the pressure.

After the initial RPV blowdown, ECCS water injected into the RPV will spill into the drywell, transporting fission products to the suppression pool and then into the primary containment.

The licensee has. conservatively assumed that the fission product release from the reactor is to mix instantaneously and homogeneously throughout the drywell. No mixing between the drywell and the wetwell is assumed for the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The licensee assumed that the initial blowdown occurs before fuel damage commences, and that the AST source terms are based on a non-mechanistic loss of ECCS flow to the reactor for two hours. After ECCS flow restoration, the rapid steaming of the ECCS liquids is assumed to quickly displace significant fractions of the airborne activity in the drywell through downcomers into the suppression chamber, providing the

Document 4: DPO Decision UNITED STAlES NUCLEAR REGULATORY COMMISSION VIASH!NGTON, 0 C, .:o50O1 February 25, 2021 MEMORANDUM TO: John G. Parillo, Senior Reactor Engineer Radiation Protection and Consequence Branch Division of Risk Assessment Office of Nuclear Reactor Regulation FROM: Raymond V. Furstenau, Director Raymond v. OiQmlf!y .signf!i by f{.i)m()IK! V.

ft.ntltl'latl Office of Nuclear Regulatory Research Furstenau ZVZ1J12.2S Q77-0S'OO-'

SUBJECT:

DIFFERING PROFESSIONAL OPINION REGARDING REGULATORY GUIDE 1.183, "ALTERNATIVE RADIOLOGICAL SOURCE TERMS FOR EVALUATING DESIGN BASIS ACCIDENTS AT NUCLEAR POWER REACTORS" (DPO 2020-002)

On August 5, 2020, in accordance with Management Directive 10.159, "The NRC Differing Professional Opinions Program," you submitted a differing professional opinion (DPO) associated with the U.S. Nuclear Regulatory Commission's (NRC's) implementation of Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors." Specifically, your DPO submittal raised concerns with the approach the staff uses to implement this guidance to assess the radiological consequences of design basis accidents.

On August 24, 2020, a DPO Ad Hoc Review Panel (the Panel) was established and tasked to meet with you, review your DPO submittal, and issue a DPO report, including conclusions and recommendations, regarding the disposition of the issues presented in your DPO. On January 19, 2021, after reviewing the applicable documents, completing internal interviews of relevant inqividuals, and completing their deliberations, the Panel issued their report to me.

The purpose of this memorandum is to respond to your DPO.

CONTACT: Nicholas DiFrancesco, RES/FO (301) 415-1115

J. Parillo Key Statement of Issues Identified by the Panel

1. Meeting intent of regulation at hand - The regulation requires an analysis of a large release to an intact containment and does not specify an accident scenario.

2. Process issue - Ongoing licensing. issues resulting from the staff proffering physical explanations for the deterministic source term.

3. Crediting safety-related systems, structures, and components (SSCs) in analyses -

Inconsistent treatment of crediting safety-related SSCs within the deterministic radiological consequence analyses.

Panel Recommendations The Panel offered the following six recommendations regarding the DPO (additional detail and bases regarding each recommendation is provided in the enclosed Panel report):

1. The agency guidance for the design-basis accident radiological consequence analysis should be revised to clarify that safety-related SSCs meeting certain criteria can, and should, be credited.

2. The agency should provide clarification to the staff when developing guidance for new radionuclide transport models to reinforce the Commission direction from the staff requirements memorandum to SECY-96-242, "Use of the NUREG-1465 Source Term at Operating Reactors."

3. The agency should to consider clarifying the Standard Review Plan (SRP) (NUREG-0800) and applicable Regulatory Guidance with regard to crediting safety-related SSCs in the radiological consequence analyses.

4. The agency should clarify guidance relating to the use of precedent licensing reviews to ensure that staff take care in applying precedence and verify that the use of past regulatory and licensing precedents remain valid given the current knowledge state and do not impose unnecessarily conservative assumptions or criteria on licensees.

5. The agency should continue to allow licensees to use RG 1.183, Revision 0, as it provides an acceptable method to address design-basis accident radiological consequences. The staff should avoid imposing the guidance contained in RG 1.183, Revision O if it would result in unnecessary conservatism that is inconsistent with regulatory requirements.

6. The agency staff should consult with the Office of Nuclear Regulatory Research to assess (1) if performance of an integrated analysis would result in a more realistic and feasible approach that avoids inconsistent analytical assumptions, and (2) if code enhancements are needed for severe accidents codes to permit overlaying of the core damage source term without the need to defeat safety-related SSCs.

In order to make a decision with regard to your DPO, I reviewed your DPO submittal and the Panel's report and I had discussions with you and the Panel. After considering all the information, I agree with the recommendations provided by the Panel. I have communicated my decision to management from the Office of Nuclear Reactor Regulation (NRR), including my position that NRR should update guidance in RG 1.183 and the SRP, consistent with the Panel report. This includes the recommendation that NRR should update guidance to allow credit for