ML17101A543

Comment (9) of Michael Tschiltz Regarding Guidance for Developing Principal Design Criteria for Non-Light Water Reactors

ML17101A543
Person / Time
Site:	Nuclear Energy Institute
Issue date:	04/04/2017
From:	Tschiltz M Nuclear Energy Institute
To:	John Monninger Rules, Announcements, and Directives Branch, Office of New Reactors
References
82FR9246 00009, DG-1330, NRC-2017-0016
Download: ML17101A543 (29)

Text

'I -

MICHAEL D. TSCHILTZ Director, New Plant, SMRs & Advanced 2t;J/ 7 Reactors 1201 F Street, NW, Suite 1100 Wa shington, DC 20004 NUCLEAR ENERGY INSTITUTE P: 202.739. 8083 mdt@nei. org nei.org R r=r~FIVE D April 4, 2017 d/31~11 Mr. John Monninger Director, Division of Safety Systems, Risk Assessment, and Advanced Reactors cf';;._/~/(, 'Jd< 4~

Office of New Reactors U.S. Nuclear Regulatory Commission Washington, DC 20555 -0001 Ei)

Subject:

NEI Comments on Draft Regulatory Guide DG-1330, " Guidance for Developing Principal Design Criteria for Non-Light Water Reactors" Project Number: 689

Dear Mr. Monninger:

On behalf of the nuclear energy industry, the Nuclear Energy Institute (NEI) 1 appreciates the opportunity to provide comments on the subject Draft Regulatory Guide DG 1330, "Guidance for Developing Principal Design Criteria for Non-Light Water Reactors." The purpose of this letter is to provide the attached comments wh ich recommend several changes to improve the clarity and completeness of DG-1330.

The design criteria provided in DG-1330 are based on a Department of Energy (DOE) effort and proposal. In addition to the comments attached, the industry recommends that NRC provide the basis for areas in which the guidance deviates from the original DOE proposal.

NRC indicates that the reason for issuing this document is to provide guidance to applicants on developing Principal Design Criteria (PDC) for non-Light Water Reactors (non-LWRs) to support the approval of construction permits, design certifications, combined licenses, standard design approvals, or manufacturing licenses. The industry believes that the proposed regulatory guidance is an important component of improving the clarity of the regulatory process and for enhancing the NRC's readiness for licensing advanced non-light water reactors .

1 The Nuclear Energy Institute ( NEI) is the organization responsible for establishing unified indust ry policy on matters affecting the nuclear energy industry, including the regulatory aspects of generic operational and technical issues. NEI 's members include all entit ies licensed to operate commercial nuclear power plants in the United States, nuclear plant designers, maj or architect/engineering firms, fuel cycle faci lities, nuclear materials licensees, and other organizations and entities involved in the nuclear energy industry .

NUCLEAR . CLEAN A IR ENERGY SUNSI Review Complete Template= ADM - 013 E-RIDS= ADM-03 )

Add= 14{ t}f/' 0fo ~

Mr. John Monninger April 4, 2017 Page 2 We appreciate the NRC staff's consideration of these comments. If you have any questions concerning this letter or the attached comments, please contact me or Thomas Zachariah (202.739.8058; txz@nei.org).

Sincerely, Michael D. Tschiltz Attachment c: Ms. Vonna L. Ordaz, NRO, NRC Ms. Deborah A. Jackson, NRO/DEIA, NRC Ms. Amy E. Cubbage, NRO/DEIA/ ARPS, NRC Mr. William D. Reckley, NRO/DEIA/ARPB, NRC NRC Document Control Desk

Affected Section Comment/Basis Recommendation

1. General NRC should clarify the language throughout Clearly state that the objective is to provide guidance to the document regarding the regulatory basis an applicant develop PDCs and not to meet the GDCs for Principal Design Criteria and the use of the as they are regulatory requirements for non-LWR regulatory gu ide once issued. reactors. This should be clear and consistent through-Principal Design Criteria (PDCs) are required out the document. For example the purpose section to be included in an application for should state:

construction permit, design certification, combined license, design approval, or This regulatory guide (RG) describes the NRC's manufacturing license. (see 10 CFR 50.35, pFepeseE! guidance on how Hie §eAeFal E!esi§A EFiteFia 52.47, 52.79, 52.137, and 52.157). E69E} iA AppeAE!ix A, " 6eAernl 9esi§A EFiteFia feF Nt1eleaF Pe*NeF PlaAts," eF =Fitle 19 eF Hie Eee!e eF 10CFR50 Appendix A States: FeE!eFal Re§t1latieAs, Paft 59 " 9emestie l::ieeAsiA§ eF PrnE!t1etieA aAE! l:::ltilii!atieA Faeilit ies" E19 EFR PaFt 59 }

The principal design criteria establish EReF. 1) appl*; te AeA li§l=lt wateF FeaeteF EAeA l::lNR) the necessary design, fabrication, E!esi§AS. =Fl=lis §t1iE!aAee may be t1see! by non-LWR construction, testing, and performance reactor designers, applicants, and licensees te may requirements for structures, systems, develop principal design criteria (PDC) for any non-LWR and components important to safety; designs, as required by the applicable NRC regulations.

that is, structures, systems, and LWR general design criteria (GDC) in Appendix A, components that provide reasonable "General Design Criteria for Nuclear Power Plants," of assurance that the facility can be Title 10 of the Code of Federal Regulations, Part 50 operated without undue risk to the " Domestic Licensing of Production and Utilization health and safety of the public. Facilities" (10 CFR Part 50) ( Ref. 1) are intended to only provide guidance to non-LWR designs. This RG derives These General Design Criteria establish Advanced Reactor Design Criteria (ARDC) from the minimum requirements for the intent of the GDC to provide more specific guidance.

principal design criteria for water- The RG also derives additiona l design-specific criteria cooled nuclear power plants similar in E!eseribes tl=le NRE's prepesee! §t1iE!aAee fer meE!ifyiA§ design and location to plants for which aAE! st1pplemeAtiA§ tl=le 69E to develop PDC that construction permits have been issued address two specific non-LWR design concepts:

by the Commission. The General sodium-cooled fast reactors (SFRs), and modular high

Affected Section Comment/Basis Recommendation Design Criteria are also considered to temperature gas-cooled reactors (mHTGRs). PDCs for be generally applicable to other types other designs can be developed using the more generic of nuclear power units and are ARDC with design-appropriate changes.

intended to provide guidance in establishing the principal design criteria for such other units.

It is industry's position, based on the above, that the GDC's of Appendix A do not establish regulatory requirements for use with non-LWR designs but provide guidance in developing and submitting PDCs with an application.

Industry believes that this RG document will essentially replace Appendix A of 10 CFR Part 50 as guidance for advanced reactors in developing PDC to be included with an application.

There are a number of statements in the draft guidance document that appear to presume the GDC in Appendix A are regulatory requirements for advanced reactors. For example, the Purpose Section of the DG states, "this regulatory guide (RG) describes the NRC's proposed guidance on how the general design criteria (GDC) in Appendix A....... .apply to non-light water reactor (non-LWR) designs." Industry believes it is unnecessary and inappropriate to attempt to make the GDC of Appendix A "apply" to non-LWRs through this guidance document but rather to simply state the objective as quidance to an applicant develop PDCs as is

Affected Section Comment/Basis Recommendation done in the second sentence of the section.

This is also consistent with the section entitled Intended Use of This Regulatory Guide in Section C.

There are a number of other places in the DG that imply conformance or alignment with Appendix A. It is recommended that a search for reference to Appendix A be performed and language appropriately clarified.

2. General In several cases, the word "reactor" is Consider removing " reactor" for consistency or explain removed from "reactor containment" in the distinction .

recognition that conta inment is a barrier between the fission products and the environment, yet " reactor containment" is retained in several other cases. (As an example, ARDC 57 and SFR-DC differ in this reqard.)

Affected Section Comment/Basis Recommendation

3. Discussion As acknowledged in the preliminary draft Without incorporating security design considerations in General guidance on non-light water reactor security the advanced reactor design criteria, add a brief Page 6 design (83 FR 13511; March 13, 2017), the discussion of the relationship and expectations for Commission's "Policy Statement on the security in design, i.e., advanced reactor design criteria Regulation of Advanced Reactors," (73 FR and security design considerations should be addressed 60612; October 14, 2008) states that the by advanced non-light water reactor developers in design of advanced reactors should " include parallel.

considerations for safety and security requirements together in the design process such that security issues (e.g ., newly identified threats of terrorist attacks) can be effectively resolved through facility design and engineered security features, and formulation of mitigation measures, with reduced reliance on human actions." NRC goes on to observe that, as we have previously commented, design considerations and associated regulatory requirements related to security are currently addressed outside of 10 CFR 50 Appendix A. We appreciate the staff's attention to distinguishing security design considerations from general design criteria. This structure should be maintained, and design considerations related to security should not be incorporated into the advanced reactor desiqn criteria .

4. Discussion, IAEA is also developing safety design criteria NRC should coordinate with mHTGR activities at IAEA in Harmonization with and safety design guidelines for mHTGRs. addition to SFRs.

International Standards, Page 10

5. Discussion, The draft regulatory guide states It is the It is recommended that this key assumption be deleted .

Kev Assumptions and responsibility of the aoo/icant to demonstrate

Affected Section Comment/Basis Recommendation Clarifications Regarding compliance with applicable severe accident the non-LWR Design and BOBE regulations and orders, Criteria, demonstrate why any that are not applicable Page 9 do not apply, and demonstrate why other design specific severe accidents or BOBE that can occur will be mitigated. "

Since ARDC/SFR-DC/mHTGR-DC apply to normal, AOOs, and design-basis events, and do not pertain to BDBE regulations, this sentence is outside the scope of this report.

6. Discussion , Seventh bullet states: "The NRC intends the Change to: " The NRC intends the ARDC to apply to the Key Assumptions and ARDC to apply to the six advanced reactor six advanced reactor technology types identified in the Clarifications Regarding technology types identified in the DOE report; DOE report; however, in some instances, one or more the non-LWR Design however, in some instances, the SFR-DC or of the criteria from the SFR-DC or mHTGR-DC may be Criteria, mHTGR-DC may be more applicable to a more applicable to a design or technology than the Page 9 design or technology than the ARDC. " ARDC."

Clarification would be useful that a "mix and match" approach is entirely appropriate - i.e.,

an entire set of criteria for a given design won 't necessarily apply.

7. Discussion, Eighth bullet states, in part: "The SFR-DC and Caveat with a statement indicating that, as with all Key Assumptions and mHTGR-DC are intended to apply to all criteria, design-specific exceptions may be proposed Clarifications Regarding designs of these technologies," which could (and defended) by the applicant.

the non-LWR Design leave the impression that the criteria in the RG Criteria, are inviolate, irrespective of specific design Page 9 attributes.

8. Appendix A The draft guidance for ARDC 16, Containment ARDC 16 language should include technology neutral ARDC 16 design, retains the original GDC language, containment requirements which can be subsequently Page A-4 thereby carrying forward design criteria applied to a specific technology. The original DOE/INL intended for a pressure-retaining light water language for ARDC 16 is provided below.

Affected Section Comment/Basis Recommendation reactor (LWR) containment. This results in limiting the applicability of the functional "Containment design.

containment concept to applicable non-LWR A reactor functional containment consisting of a designs, and appears to be inconsistent with structure surrounding the reactor and its cooling system the Commission's position on alternatives to a or multiple barriers internal and/or external to the leak tight containment, as discussed in SECY reactor and its cooling system/ shall be provided to 93-092 and the associated SRM. control the release of radioactivity to the environment and to assure that the functional containment design Advanced reactor containment design conditions important to safety are not exceeded for as guidance should flow logically from ARDC 16 long as postulated accident conditions require. "

to the SFR and mHTGR design criteria. ARDC 16 should be a high-level technology-neutral The concept of a functional containment would be of design criteria from which technology-specific interest for application to other technologies. Applying design criteria are derived. this recommendation would provide a high-level technology-neutral ARDC which could be used to obtain Commission approval of containment performance criteria. SFR and mHTGR DC 16 would then serve to illustrate how technology-specific design criteria can be derived from ARDC 16.

9. Appendix A Clarify that use of ARDC 16 [per industry Revise rationale to state, " ...However, it is also ARDC 16 comment##] for non-LWR designs other than recognized that characteristics of the coolants, fuels, Page A-4 mHTGRs may "be subject to a policy and containments to be used in other non-LWR designs decision .. ." Making a justification, similar to could share common features with SFRs and that for research reactors and non-power mHTGRs .. .Use of tfte ARDC 16 for non-LWR designs reactors has basis in NRC policy and should other than mHTGRs DC 16 will may be subject to a not require a Commission-level policy decision. policy decision by the Commission. If a reactor is able to demonstrate safety margins and/ or consequences on Discussions of Commission policy decisions on the order of those demonstrated by non-power and functional containment need to be worded research reactors, a functiona l containment may be carefully. For the modular HTGR, a policy justified, and the reactor may be able t o use ARDC 16 decision is not needed regarding the general without a Commission level policy decision . 5ee acceptability of applying a functional rationale for rnHTGR DC 16 for fu rther information on containment (radionuclide retention) approach " ' ' -

, .... y

..J ......... *

• II

Affected Section Comment/Basis Recommendation that differs from a conventional LWR high-pressure, low-leakage structure.

However, based on the SRM to SECY-03-0047, a policy decision is needed regarding the performance criteria to be applied to a functional containment. The information located in the mHTGR-DC 16 rationale correctly states that a policy decision regarding functional containment performance requirements and criteria will be needed. It's noted that containment performance criteria for LWRs are provided in 10 CFR 50 Appendix J, rather than in the GDC of Appendix A.

10 Appendix A, Clarify "A reliable power system is required for Modify to: "A reliable power system is required for ARDC 17, SSCs during postulated accident conditions" to SSCs during postulated accident conditions when those Page A-4 apply to SSCs whose safety performance relies SSCs' safety functions require electric power."

on electric power 11 Appendix A The following text is confusing: "The existing Suggest rewording to: "The single switchyard ARDC 17 single switchyard allowance remains available allowance under GDC 17 is not eliminated because of Page A-4 under ARDC 17. If a particular advanced the changes in ARDC 17; if a particular advanced design requires the use of GDC design .. .

single switchyard allowance wording, the designer should look to GDC 17 for guidance when developinq PDC."

12 Appendix A ARDC 17 states the safety function for the Revise ARDC 17 with respect to the postulated accident ARDC 17 electrical systems "shall be to provide safety function, or clarify the scope of "vital functions" ARDC 26 sufficient capacity, capability, and reliability to with the Rationale.

ensure that.. .vital functions that rely on electric power are maintained in the event of Revise the Rationale discussion on applicability of postulated accidents." The scope of "vital ARDC 17 to address the use of electrical power for the functions" is unclear. For example, it is unclear performance of the prescribed safety functions.

if the independent and diverse means of

Affected Section Comment/Basis Recommendation shutdown prescribed by ARDC 26 paragraph 2 is considered such a vital function .

Further, the Rationale for ARDC 17 states "If electrical power is not required to permit functioning of SSCs important to safety, the requirements in the ARDC are not applicable to the design. In this case, the functionality of SSCs important to safety must be fully evaluated and documented in the design bases." The requirements of ARDC 17 are related to performance of the prescribed safety functions (e.g., sufficient redundancy "to perform their safety functions").

Accordingly, it appears the appropriate test for applicability of ARDC 17 is whether electrical power is required to perform the specifically prescribed safety functions, not the functioning of SSCs important to safety more generally.

13 Appendix A, This criterion presumes that operator action is Consideration should be given to an applicant ARDC 19, required and that operator actions, including demonstrating that operator action, including Page A-6 monitoring, must be performed from a single monitoring, is not required for safety, and/or that any location (i.e., a control room). necessary actions, including monitoring, could be demonstrated to be feasible from additional and/or redundant and/or remote locations.

Affected Section Comment/Basis Recommendation 14 Appendix A, The way the text is written still appears to As with some other sections, frame with "As applicable ARDC 19, assume some fundamental, legacy needs in a to plant design:"

Page A-6 power plant. None of this makes sense if operators have literally zero ability to influence the safety of the plant because it is physically inherent (note: not to be confused with "inherent" safety as defined by the IAEA, which requires no decay heat) 15 Appendix A, It appears assumed that control/protection As with some other sections, frame with "As applicable ARDC 25 through 28, systems are required for reactivity control. It to plant design:"

Page A-7 also assumes that the ultimate reactivity protection mechanism is still an active function. This assumption is not necessarily true for all designs. The term "system" indicates active/desiqned to us.

16 Appendix A, (1) Capability (1) is specific to having a means Define "Appropriate Margin" AND ARDC 26, to shut down the reactor in regularly Page A-7 occurring situations. The move from Change wording to the below (italics indicates changed specified acceptable fuel design limits to wording, bold indicates added wording) fission product barriers is a significant improvement towards technology Reactivity control systems shall include the following neutrality, enabling accurate safety capabilities:

assessment of both more conventional fuel (1) A means of shutting down the reactor shall be forms with more complex fuel forms provided to ensure that, under conditions of normal including liquid fuel forms on the same operation, including anticipated operational basis. occurrences, and with appropriate margin for malfunctions, design limits for safety-related fission That being said, there was concern that there product barriers are not exceeded.

are some possible components considered as (2) A means of shutting down the reactor and fission product barriers could fail without maintaining a safe shutdown in anticipated operational significant impact to safety. Therefore words occurrences and postulated accidents, with appropriate were added to ensure that the focus is on only margin for malfunctions, shall be provided. If the those fission product barriers that are safety-

Recommendation Affected Section Comment/Basis primary means for shutdown is not inherent, passive, or related. shown to have a probability of failure an order of magnitude less than that of postulated accidents, a (2) Many industry comments included second means of reactivity control shall be provided reasoning that two independent means for that is independent, diverse, and capable of achieving shutting down the reactor and maintaining and maintaining safe shutdown both for anticipated shutdown may not be needed, especially operational occurrences and postulated accidents.

for reactor types that have natural or (3) A system for holding the reactor subcritical in the passive means for shutdown as the primary means. In addition, the long term or in an equilibrium condition naturally achieved by the design under cold conditions shall be requirement for two fully independent means both capable of achieving and provided.

maintaining shutdown does not seem to be the standard for LWRs.

This presents the simplest wording that allows for reactors with inherent or passive shutdown fundamental to the physics of the system to make a justification that a second means would be superfluous. It also allows for reactors to make a probability risk assessment to make a similar justification.

The wording change from "design basis events" to "anticipated operational occurrences and postulated accidents" is taken from the NRC's Rationale and ensures that what is being referred to is clearly outlined terminology in the regulation.

(3) The requirement of subcriticality may not be the most appropriate measure of safe shutdown. For example, it has been demonstrated in various reactor tvoes that

Affected Section Comment/Basis Recommendation a safe, long term shutdown could be achieved naturally without rods or coolant even if brief moments of criticality occurred. (see "Secondary shutdown systems of Nuclear Power Plants," ORNL-NSIC-7, January 1966). Wording was taken directly from the NRC Rationale to expand the capability to account for such a capability in certain designs.

With the addition of the phrase " appropriate margin for malfunctions, " it is important that the subjective phrase be defined by NRC.

This wording is an attempt to define "appropriate margin " with options for both deterministic and risk-informed scenarios for malfunction. Depending on the reactor type, it may be preferred to utilize the simplicity of a deterministic approach. There also may or may not be enough data to utilize a risk-informed approach. For others, a risk-informed approach may more accurately determine appropriate margin.

The previous metric of maintaining fission product barriers is kept as the primary metric in this measurement of margin. The definition could be:

(1) A single active failure must not result in exceeding design limits for safety-related fission product barriers, or

Affected Section Comment/Basis Recommendation (2) The probability for a malfunction of the means must not be greater than the frequency for AOOs. If the probability is greater than the frequency for postulated accidents by an order magnitude or more, that malfunction must not result in exceeding design limits for safety-related fission product barriers.

17 Appendix A, The second to last paragraph of the ARDC 26 Recommend restating the rational to say ARDC 26, rationale states Page A-7 "The second sentence of ARDC 26(2) refers to "The second sentence of ARDC 26(2) refers to a ffleaflS a means of achieving and maintaining ofacllie~'ifl[j Dfl<i fflD/fltaiRifl[j shtJtdOWA that is shutdown that is im/l,ortant to safetx_ but if]1fJeffaflt te salefi.' ef:ft Ret fiecessDFl'/j' safefi.' re/ate/.

not necessarilr.. safetr.. related. The second The seceR/ means of reactivity control which serves as means of reactivity control serves as a backup a backup to the salety related primary means an~ as to the safety-related means an~ as such/ such/ margins for malfunctions are not required but the margins for malfunctions are not required but second means shall be highly reliable and robust (e.g./

the second means shall be highly reliable and meet ARDC 1 -5). "

robust (e.g./ meet ARDC 1 -5)."

The distinction between the terms "important to safety" and "safety-related" is not properly defined. To avoid confusion, the statement should be revised.

Affected Section Comment/Basis Recommendation 18 Appendix A ARDC 35 states ':4 system to provide sufficient It appears that the cited ARDC 35 text expands the ARDC 35 emergency core cooling shall be provided. The scope of the existing GDC, and is therefore outside of A-14 system safety function shall be to transfer the scope of this ARDC effort. Absent further heat from the reactor core such that effective information regarding the intent of these words, it is core cooling is maintained and fuel damage recommended that they be deleted from the criterion.

is limited. "

Regarding the addition of the words and fuel damage is limited"to the first paragraph of the criterion, the rationale does not provide guidance for how these new words (which reflect an expansion relative to GDC 35) should be interpreted or why they have been added.

The added words are ambiguous when considering (1) to what level should fuel damage be limited? (2) What are the appropriate measures of fuel damage? (3)

How would fuel damage be interpreted for a molten salt reactor or for a modular HTGR?

19 Appendix A Clarify "important" refers to "important to Change to "The structural and equipment cooling ARDC 45 safety" systems shall be designed to permit appropriate A-19 periodic inspection of important safety related components, such as heat exchangers and piping, to ensure the inteqritv and caoabilitv of the systems."

20 Appendix A Clarify applicability to SSCs with a safety Change to "=tfle structural Safety Related structural and ARDC 46 function equipment cooling systems shall be designed ... "

A-19 21 Appendix A Editorial: "The example at the end of subpart As indicated ARDC 50 1 of ij:ie ARB GDC 50 is LWR specific. .. "

A-20

Affected Section Comment/Basis Recommendation 22 Appendix B In several cases, SFR-DCs indicate "same as Consider indication, where applicable, that only General ARDC." Some others do not indicate this, difference from ARDC is coolant boundary designation .

when the only change is from "reactor coolant boundary" to "primary coolant boundary."

23 Appendix B In many cases, the SFR-DC rationale include: Replace "implies" with "indicates" for consistency.

General "The use of the term "primary" indicates that the SFR-DC are applicable only to the primary cooling system, not the intermediate cooling system." In several instances, however, "indicates" is replaced with "implies," which connotes less certainty as to applicability.

24 Appendix B As regard quality standards and records, and It is suggested to add that design codes adapted to SFR-DC 1 and 10 reactor design, no specific SFR criteria are SFR specificities (high temperature.. .) must be defined'~

proposed 25 Appendix B The definition of the primary coolant boundary It is therefore proposed to state that " Each part of the SFR-DC 14 includes the cover gas boundary. Therefore, primary coolant boundary shall be designed, fabricated, the Criterion 14 requiring an extremely low erected, and tested so as to have aA exfFefflel';' low probability of abnormal leakage for cover gas probaeility a prevention level of abnormal leakage, of leakage is not necessary. A cover gas leakage rapidly propagating failure, and of gross rupture, would lead to very limited safety commensurate with the consequences of such failures'~

consequences (no impact on the fission process, no impact or limited radiological consequences). This allows for safety valves on the cover gas system to limit abnormal pressure on the reactor vessel. On the other hand, the failure of the reactor vessel could have very severe consequences (e.g. reactivity insertion, failure of the core coolability).

26 Appendix B It is indicated that the reactor containment is It is therefore proposed to modify the first sentence of SFR-DC 16 a pressure retaining structure surrounding the the criterion as: .4 reactor containment consisting of a B-5 reactor and its cooling systems. In case of high strength, low leakage, presStJFe reta/Aiff!J structure SFR, it is possible to limit the pressure surroundino the reactor;:;, .:: 't:: __ ,,. .. ., :;*.::. . --..::shall be

Affected Section Comment/Basis Recommendation loadings on the containment structure in provided to control the release of radioactivity to the accident conditions. For example the rooms environment and to assure that the reactor with sodium circuits can be designed so that containment design conditions important to safety are the effect of a sodium leak or fire would not not exceeded for as long as postulated accident result in significant pressure on the conditions require. "

containment structure and the pressure effect could be limited to the room where the leak Additionally, remove the phrase "and its primary cooling occurs. Also, the reactor cooling systems could system."

include secondary cooling systems which are partially outside the containment structure where this can be particular concern is cooling systems with air as the heat sink, for which sodium/air heat exchanger must be placed outside of the containment.

27 Appendix B Under rationale, statement that "all past, Delete "and planned" SFR-DC 16 current, and planned SFR designs use a high-B-5 strength, low-leakage, pressure-retaining containment concept" seems broader than can be substantiated without knowledge of illl planned desiqns.

28 Appendix B Editorial : "The existing single switchyard As indicated . However, also refer to comment on ARDC SFR-DC 17 allowance remains available under ARBt SFR- 17 suggesting rewording of this rationale discussion.

B-6 DC 17 ... "

29 Appendix B GDC 26 and GDC 27 requirements are: Recommend retaining GDC 26 and 27 unchanged as SFR-DC 26

• Two independent reactivity control SFR-DC 26 and SFR-DC 27. GDC 26 and 27 are SFR-DC 27 systems of different design applicable for currently licensed and operating LWRs.

principles shall be provided. The reactivity control requirements currently in place

• One of the systems shall use control for LWRs are sufficient for SFRs.

elements and be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences (AOOs), and

Affected Section Comment/Basis Recommendation with appropriate margin for malfunctions such as stuck control elements, specified acceptable fuel design limits are not exceeded .

• The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes to assure acceptable fuel design limits are not exceeded .

• One of the systems shall be capable of holding the reactor core subcritical under cold conditions.

• The reactivity control systems shall be designed to have a combined capability of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck control elements the capability to cool the core is maintained .

Current BWRs and PWRs in the US have two independent systems for controlling reactivity through movement and positioning of control rods.

To attain the desired core power level and power distribution during normal operation, one reactivity control system is used to position control rods to compensate for reactivity due to changes in temperature and fuel burnup. BWRs also used core flow

Affected Section Comment/Basis Recommendation and PWRs also use boration to help control reactivity during normal operation. To ensure all safety criteria are met during AOOs and DBAs, a second reactivity control system is used to provide rapid, full insertion of all control rods (scram). The circuitry and hardware used to move the control rods are completely independent for the two reactivity control systems.

The reactivity worth of the control rods is sufficient to ensure reactor shutdown when the rods are fully inserted by either control system for BWRs. For PWRs, control rod insertion and boration ensure reactor shutdown.

US LWRs have implemented design features to provide an alternate method for reactor shutdown in the event that the reactivity shutdown system (scram) fails .

For PWRs, alternate control rod insertion methods in the event of scram failure have been implemented (same control rods as normal scram, but an independent method

• for inserting the rods) . For BWRs, standby liquid boron injection systems are used to provide an alternate method for reactor shutdown. These alternate means to shut down the reactor are required to meet 10CFRS0.62 requirements. Note, these I

alternate means of shutdown are for a beyond desiqn basis event and the I

I I

_J

Affected Section Comment/Basis Recommendation requirements are not addressed in the GDC.

Requirement differences with NRC SFR-DC 26 :

• Item (1) of SFR-DC 26 changes

" specified acceptable fuel limits" to

" design limits for fission product barriers". Challenges to primary coolant boundary or containment boundary are addressed in other GDCs. Change is not necessary, but does not add new requirement.

• Item (2) of SFR-DC 26 changes the requirement to "provide capability to cool the core" during " postulated accidents" to "maintaining a safe shutdown under design basis events". The reactivity control system requirement has been extended from ensuring core damage does not prevent core cooling to including other aspects (e.g. heat removal from primary system) of safe shutdown.

Additional requirements to achieve safe shutdown are addressed by other GDCs.

The term "design basis events" is not used in the GDCs.

• Item (2) of SFR-DC 26 adds the requirement to have a second independent shutdown system for design basis events. 10CFR does not require a second independent shutdown system for design basis events. 10CFR requires an alternate

Affected Section Comment/Basis Recommendation means of shutdown for beyond design basis events (10CFRS0.62).

• SFR-DC 26 eliminates the requirement that the reactivity control system for normal operation reactivity control be independent from the reactivity control system used for shutdown (scram).

30 Appendix B Similar comment as the one for SFR-DC 14. It is therefore proposed to state that "Each components SFR-DC 30 The definition of the primary coolant boundary that -aFe is parts of the primary coolant boundary shall includes the cover gas boundary. A cover gas be designe~ fabricate~ erected and tested .fe-tfle leakage would lead to very limited safety highest q&a/ity staruiards practical with high quality consequences (no impact on the fission standards, consistent with its safety_ significance '~

process, no impact or limited radiological consequences). This allows for safety valves on the cover gas system to limit abnormal pressure on the reactor vessel. On the other hand, the failure of the reactor vessel could have very severe consequences (e.g. reactivity insertion, failure of the core coolability).

31 Appendix B The goal of GDC 33 is that the cooling Replace the phrase "specified acceptable fuel design SFR-DC 33 function of the primary heat removal system limits are not exceeded" with the phrase "the cooling shall not be impacted during normal operation functions of the primary heat removal system and the by primary coolant inventory loss due to residual heat removal system are not impacted".

leakage from the primary coolant boundary To eliminate redundancy, delete the phrase "for and rupture of small piping or other small protection against small breaks in the primary coolant components which are part of the boundary. boundary".

For SFRs specifically, the primary concern is ensuring primary coolant inventory is sufficient to maintain the cooling function for the primary heat removal system. This ensures specified acceptable fuel design limits are not exceeded.

32 Appendix B SFR-DC 34 deleted reference to postulated Explain the reason inq for SFR-DC 34 beinq for normal

Affected Section Comment/Basis Recommendation SFR-DC 34 accidents (e.g. DBAs) without an explanation operations and AOOs, similar to the explanation in the rationale section. provided for SFR-DC 35.

33 Appendix B For SFRs, the residual heat removal Replace the first paragraph of SFR-DC 35 with the SFR-DC 35 system may be all that is required to following paragraph:

provide adequate heat removal during * "A system to assure sufficient core cooling during postulated accidents. postulated accidents and to remove residual heat following postulated accidents shall be provided.

SFR-DC 34 is specified as being applicable The system safety function shall be to transfer for normal and AOO conditions. However, heat from the reactor core during and following residual heat removal will also be necessary postulated accidents such that fuel and clad for postulated accident conditions and damage that could interfere with continued should be addressed in SFR-DC 35. effective core cooling is prevented and the design conditions of the primary system boundary are not The draft SFR-DC 35 added "and fuel exceeded."

damage is limited". Other than maintaining effective core cooling, the meaning of this statement is not clear - what is being prevented by limiting the fuel damage?

Suggest using wording similar to that used in GDC 35; that is use" ... . such that fuel and clad damage that could interfere with continued effective core cooling is prevented .... " instead of " .... such that effective core cooling is maintained and fuel damage is limited ... ".

SFR-DC 35 does not address protection of the primary coolant system boundary. Add

" .. .and the design conditions of the primary system boundary are not exceeded."

34 Appendix B The title of these SFR-DC refers to the Revise title of SFR-DC 36 to Inspection of emergency SFR-DC 36 & 37 " residual heat removal system". The text that core cooling system.

follows refers to the emergency core cooling

Affected Section Comment/Basis Recommendation system. While a single system may be Revise title of SFR-DC 37 to Inspection of emergency provided to perform both residual heat core cooling system.

removal and emergency core cooling functions, it would be logical for the title and the text to use the same nomenclature to describe the system.

35 Appendix B The opening sentence is confusing. The opening sentence needs to be revised to make its SFR-DC 44 meaninq clearer.

36 Appendix B SFR structures are sensitive to pressure and it We propose to state that "the reactor containment SFR-DC 52 may be chosen to avoid high pressure structure and other equipment that may be subjected elevation in the containment design during to containment test conditions shall be designed so that leakage rate testing, in order to preserve the periodic integrated leakage rate testing can be facility and prevent undesirable over or under conducted to demonstrate resistance at containment pressurization risks during those tests. It may design pressure '~

be chosen to perform those tests at a pressure below the containment design pressure, in order to extrapolate them at the containment design pressure (in this case the relevance of the extrapolation will of course have to be justified).

37 Appendix B As indicated in criterion 57, an isolation of To ensure coherency of the text, this could be reflected SFR-DC 54 lines penetrating the reactor containment in the Criterion 54: Piping systems penetrating the structure may not be required in some cases. reactor containment structure shall be provided with This could for example could apply to the leak detection isolation if necessary and containment intermediate heat transport system capabilities(..)"

penetrating the reactor containment (provided adequate justification is given).

38 Appendix B Why is "Isolation valves outside Add the wording to SFR-DC 56.

SFR-DC 56 containment. .. " deleted? It's not deleted in 55.

It appears from the wording that the intent was that this phrase NOT be deleted from SFR-DC 56. Deletion may have been

Affected Section Comment/Basis Recommendation unintentional.

39 Appendix B The first sentence, "If an intermediate coolant Rewrite the DC to state "If an intermediate cooling SFR-DC 70 system is provided, then the system shall be system is provided, then the system shall be designed If designed to transport heat from the primary with sufficient margin ...

coolant system to the energy conversion system as required," is not required .

40 Appendix B Sodium freezing may not impact the safety Add phrase " ... .if necessary to ensure that the safety SFR-DC 72 function of all systems. function of the system is accomplished " to the beginning of the first sentence.

41 Appendix B "Heating systems shall be provided for To minimize confusion, restate as : "Heating systems SFR-DC 72 systems and components important to safety, shall be provided for systems and components that are which contain or could be required to contain important to safety, which and that contain or could be sodium." could be inferred to mean that all required to contain sodium."

systems and components important to safety contain or could be required to contain sodium.

42 Appendix B Is the intent of the last sentence to ensure Recommend deleting the last sentence.

SFR-DC 73 that all sodium systems be in inerted enclosures or guard vessels? Not all plant systems containing sodium need to be in inerted spaces.

43 Appendix B " Special features, such as inerted enclosures Replace this sentence in its entirety with : "Systems SFR-DC 73 or guard vessels, shall be provided for systems from which sodium leakage constitutes a significant containing sodium." implies a significant safety hazard shall include measures for protection, hazard exists for any system containing such as inerted enclosures or guard vessels. "

sodium .

44 Appendix B Fire protection and mitigation due to sodium- Delete phase " ... , including mitigation of the effects of SFR-DC 74 water interaction is covered by SFR-DC 3 and any resulting fire involving sodium. "

SFR-DC 73.

45 Appendix B SFR-DC 70 states "The intermediate coolant Recommend deletion of SFR-DC 75, 76, and 77.

SFR-DC 75 system to be designed with sufficient margin SFR-DC 76 to assure that (1) the design conditions of its

Affected Section Comment/Basis Recommendation SFR-DC 77 boundary are not exceeded during normal If SFR-DC 76 is not deleted, it should include wording operations and anticipated operational such as "commensurate with their importance to occurrences, and (2) the integrity of the safety."

primary coolant boundary is maintained during intermediate coolant system accidents. "

SFR-DC 75, 76, and 77 are superfluous when evaluated in combination with the cited text from SFR-DC 70. SFR-DC 75, 76, and 77 appear to be applicable when the role of the intermediate coolant system is commensurate with a safety function. However, other than the case when it could serve as a path for decay heat removal, the intermed iate coolant system does not have any safety function.

If the intermediate cooling system provides a safety-related heat removal capability, then SFR-DC 34-37 and SFR-DC 78 specify its requirements. The quality and fracture prevention requ irements specified in SFR-DC 75 and 76 are supplementary requirements that are not consistent with the requirements for the decay heat removal and emergency core cooling systems specified in SFR-DC 34 and 35. Likewise, the inspection and testing

Affected Section Comment/Basis Recommendation requirements specified in SFR-DC 77 for the intermediate cooling system are contained in SFR-DC 36 and 37. Therefore, for the case where the intermediate cooling system provides safety-related heat removal capability, SFR-DC 75, 76, and 77 are redundant and unnecessary.

If the intermediate cooling system does not provide safety-related heat removal capability, then only the requirements of SFR-DC 70 are necessary to specify the system design with appropriate margin to assure the design conditions of its boundary and the integrity of the primary coolant boundary. Therefore, for the case where the intermediate cooling system does not provide safety-related heat removal capability, SFR-DC 75, 76, and 77 are also redundant and unnecessary.

46 Appendix B It is possible that there either be such a Move the first sentence to the end with added wording SFR-DC 78 configuration or that there be not be enough described below.

liquid metal to cause a severe consequence or even a significant consequence due to reactions with either air or water or both, both After " compatible" in the second sentence, add "or in terms of the reaction itself as well as

Affected Section Comment/Basis Recommendation consequence to the reactor and safety system incompatible".

functions.

Instead of being prescriptive, there needs to Add wording to the end to read: "If the primary coolant be a mechanistic method to determine system interfaces with a structure, system, or whether multiple boundaries are necessary.

component containing fluid that is chemically Ultimately, the prescriptive condition for two boundaries is redundant; for both fluids and incompatible with the primary coolant, and cannot meet coolants which are compatible or condition (1) and condition (2), the interface location shall be designed to ensure that the primary coolant is incompatible, the required conditions should separated from the chemically incompatible fluid by two be the same, which are the conditions (1) and redundant, passive barriers.

(2).

So long as there is no failure of the intended safety functions of structures, systems or components important to safety or result in exceeding the fuel design limits, then the size of the reaction is small enough to justify not needing redundant boundaries.

47 Appendix B The requirement to ensure that "primary Delete SFR-DC 79 SFR-DC 79 coolant sodium limits" are not exceeded as a result of cover gas leakage are already addressed in SFR-DC 71, item (4).

48 Appendix C General Many of the proposed mHTGR GDC retain the The single failure requirement should be replaced with statement "assuming a single failure". This a probabilistic (reliability) criterion.

inclusion makes no reference to SECY-03-0047 and the Commission SRM that described the replacement of the sinole failure criterion with

Affected Section Comment/Basis Recommendation a probabilistic (reliability) criterion.

49 Appendix C The requirements as written imply the primary De-emphasize the pressure retention function of the mHTGR-DC 14, 30, 31, helium pressure retention is a safety function helium pressure boundary.

32 similar to LWRs.

mHTGR-DC 70 correctly emphasizes seismic stability However, it is important to note that although and geometric stability of the reactor vessel system .

the leak tightness and high quality of the helium pressure boundary is necessary for However, emphasis on T/H properties of the reactor commercial operation of mHTGRs, the vessel at uninsulated the core region is lacking.

pressure retaining function of the helium pressure boundary is not a required safety function.

The safety function of the reactor vessel and its support system is to maintain core coolable geometry and provide sufficient conduction and convection heat transfer properties in the core reqion.

50 Appendix C The addition of "heat removal systems" Clarify the role of the RCCS for heat removal under mHTGR-DC 15 appears to be limited solely to connected normal operations and AOOs .

systems, i.e., the steam generator.

Clarification is needed as to the role of the RCCS for heat removal under normal operations and AOOs.

51 Appendix C Editorial: "The existing single switchyard As indicated. However, also refer to comment on ARDC mHTGR-DC 17 allowance remains available under AA-9 17 suggesting rewording of this rationale discussion.

mHTGR-DC 17 .. ."

52 Appendix C Delete "as defined in § 50.2" as this is implicit Delete "as defined in § 50.2" mHTGR-DC 19 in all of the GDC statements.

53 Appendix C The existing GDC includes the wording Recommend establishing consistency between mHTGR-mHTGR-DC 26 "specified acceptable fuel design limits", while DC 26 and other design criteria mentioned.

the proposed mHTGR-DC does not include the

Affected Section Comment/Basis Recommendation replacement "specified acceptable system radionuclide release design limits" wording .

The wording that "design limits for fission product barriers are not exceeded" is imprecise and moves the intent from maintaining fuel design limits to fission product barriers. The rationale desribes:

"Additionally, "specified acceptable fuel design limits" is replaced with "design limits for fission product barriers" to be consistent with the AOO acceptance criteria." This appears to be inconsistent with other design criteria which include SARRDL.s. See proposed mHTGR-DC 10, 17, 20 and 25.

54 Appendix C With the inclusion of AOOs within mHTGR Delete mHTGR-DC 29 mHTGR-DC 29 GDC 20, 25, and 26, it is recommended that this GDC is duplicative and can be deleted.

55 Appendix C The word "passive" implies that only a passive Remove the word " passive" mHTGR-DC 34 system is to be provided. Maintaining mHTGR-DC 71 geometry is needed for both active and mHTGR-DC 72 passive means of heat removal.

Note that proposed new mHTGR-DC 72 does not mention passive (while the rationale does).

56 Appendix C Add the word "system" after residual heat Add the word "system" after residual heat removal.

mHTGR-DC 36 removal.