ML20215G030

Discusses Modular HTGR Review Schedule,Per 870527 & 28 Meetings W/Doe,Ornl & Bnl.Based on Status & Experience W/ Review,Unclear Whether Scheduled Dates for Development of SER Can Be Met.List of Attendees & Meeting Agendas Encl

ML20215G030
Person / Time
Issue date:	06/16/1987
From:	Morris B NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:	Gavigan F ENERGY, DEPT. OF
References
PROJECT-672A NUDOCS 8706230097
Download: ML20215G030 (17)

Text

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o 1 0 1987 Mr. Francis X. Gavigan, Director Office of Advanced Reactor Programs Office of Nuclear Energy U. S. Department of Energy i

Washington, DC 20545

Dear Mr. Gavigan:

On May 27 and 28, 1987, members of the NRC staff and its contractors from ORNL and BNL met with representatives of DOE and its contractors to review the following(chapters from the Modular HTGR Preliminary Safety Information Document PSID), Project 672:

Chapter 6, " Buildings and Structures," Chapter 9, " Service Systems," Chapter 10 " Steam and Energy Conversion Systems, Chapter ll, " Occupational Radiation Protection," and Chapter 13. " Conduct of Operations." The agenda and list of atter.uaes are given in Enclosures 1 and 2, respectively. Certain asterisked items on the agenda were not discussed at the May meeting but are being deferred to the forthcoming meeting on June 18, 19, 1987. The Action Items and Clarifications (including requests for additional information) resulting from this meeting are given in Enclosure 3 for your action. Your response to these items is required by July 10, 1987, in order for us to maintain our current review schedule.

I would like to take this opportunity to comment regarding the MHTGR review schedule as discussed with you on June 9, 1987. As you recall, at the start of the MHTGR review we mutually developed a review schedule consisting of monthly meetings with DOE and its contractors from January, 1987 until July, 1987 to review each chapter of the PSID. This series of review meetings was intended to provide the staff an opportunity to ask questions on the MHTGR design, request additional information and address many of the key technical issues at the appropriate PSID chapter review meeting. Timely response to staff comments and requests for additional information was necessary to maintain the schedule (see letter T. Speis to F. Gavigan, dated October 17, 1986) and to allow the staff to work on resolution of many of the key issues over the course of the review.

In addition, timely staff review and consensus on the key technical issues was also necessary to ensure our Safety Evaluation Report (SER) was developed on schedule (draft by 9/30/87 and final by 1/31/88).

It was recognized at the time the review schedule was developed that it was an ambitious success oriented plan, considering the innovative nature of the MHTGR design, the unique technical and policy issues raised by such a design and the desire of the Department to have the review completed as soon as possible. To date both the Department and its contractors and the staff and its contractors have been working diligently to complete the review according to the plan and schedule originally established.

However, based on the status and our experience with the review to date, it is not clear whether the scheduled dates for development of the SER can be met. The following two factors contribute to this conclusion:

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k 8706230097 870616 PDR PROJ 672A PDR

1 JUN 161987

_2 (1) The MHTGR review schedule is ahead of the other advanced reactor reviews and, therefore, is taking the lead in dealing with certain key technical and policy issues resulting from the novel design approaches being proposed. As such, the staff may require more time for satisfactory.

resolution of these issues, recognizing that important precedents may be established (for example, our response to your' proposed top level regulatory criteria has taken longer than originally planned). Our success in this regard is, of course, also dependent upon the timeliness and quality of.your response to staff comments and requests for additional information in these areas.

-(2) Previous and potential delays in submittals of documentation or in responding to staff requests are causing a large volume of material to be submitted at the end of the review. period.

Specifically,'an earlier -

delay involved the PRA submittal, and current delays involve your responses on "important to safety classification," the uncertainty and sensitivity study on inherent reactivity feedback characteristics, j

deferral of discussion of agenda items on source terms and radionuclide

' control, and your response on document control procedures being applied to the MHTGR-PSID. Your records will indicate the specific due dates for 4

these responses, but the'important issue is that the review schedule has been carefully structured to allow a balanced workload and the NRC staff cannot absorb submittal of a large quantity of material at a late stage:

in the process.

As discussed with you on June 9, 1987 we will work, as best as possible, to maintain the review schedule. However, in view of the_above items I suggest that upon completion-of our scheduled series of meetings in July 1987,.upon receipt of the information requested as a result of our April 1987, May 1987 and remaining meetings, and upon receipt of the delayed and deferred. items.

listed above, we meet to discuss any changes in our expected completion date for the SER and ways of minimizing such changes.

If you have any questions please do not hesitate to contact Dr. Peter Williams,l the-Project Manager-for Project 672.

Sincerely, 1j Bill H._ Morris, Director 1

Division of Regulatory Applications Office of Nuclear Regulatory Research cc: D.~Mears, GCRA A. Millunzi, DOE-

Enclosures:

1. Agenda-J 2.' List of Meeting Attendees 1

3. Action Items and Clarifications OFC :ARGIB:DRA

ARGIB:DRA

.:ARGIB:DRA

DD:DRA

D:DRA NAME :PWilliams:rr JNWilson;

TKing

ZRosztoczy- :BMorr s 1

q SATE':6/ /87'

6/

/87'

6/- '/87-

6/

/87-

6/// /87 1

. ro JUF 1 : 1987 DISTRIBUTION RES. Circ

.Chron

'ARGIB R/F E. Beckjord T. Speis B. Morris.

Z. Rosztoczy.

T. King J. N. Wilson C. Allen

.R. Landry P. Williams-EJ Flack ~

'R. Baer N. Anderson

-F.

Cherny.-

S. Shaukat.

R. Johnson D. Thatcher J. Hulman J.-Glynn L. Soffer J. Read D. Cleary A. Murphy G. Arndt.

R. Kirkwood M..Lamastra-R. Erickson B. Mendlesohn H. VanderMolen E.~ Chelliah L. Beltracchi:

('

'F. Congel' J.- Swift

0. Lynch D. Matthews'

.R. Senseney M. Spangler

,F.

Coffman S. Ball, ORNL P. Kroeger, BNL

.G. VanTuyle, BNL' R. Ireland,. Reg. IV-M. El-Zeftawy,:-ACRS/H-1026

PDR-Project 67.2..NhAT3 PhijFet'3 File 7672C(C Files)go g.

+

+-

i

t Attendees NRC/ DOE Meeting on MHTGR May 27 and 28, 1987 4

P. M. Williams **

NRC/RES/ARGIB 492-9613-Ralph'Landry NRC/RES/ARGIB 492-4914 Thomas L. King NRC/RES/ARGIB-492-7014

)

Richard E. Johnson

• NRC/RES Peter G. Kroeger BNL.

(516) 282-2610 Sunil K.~Ghose*

Bechtel 1

George Gurnis Stone'& Webster (617) 589-7403

Roger Kenneally*

NRC/RES i

John M. Oddo Stone & Webster

-(617) 589-7403 i

Gunter Arndt*

Malcolm LaBar GCRA' (619)'455-9500 John Recknagel**

PSE&G (201) 530-8110 Lloyd P. Walker

.SWEC/GCRA (619) 455-9500 J. W._Kendall GCRA' (619) 455-9500 B. T. Mendelsoln.

NRC/RES/RSGB

_492-9631 Donald Graf PDC 0 (619) 455-4294 A. C; Millunzi DOE (301) 353-3405 Tony Nylan GA (619) 455-2580-J. C. Cunliffe Bechtel (415)_768-2227 Dan Hears **

GCRA (619) 455-9500 Syd Ball ORNL (615).574-0415.

David L. Moses ORNL (NRC Programs)

(615) 574-6103/

FTS 624-6103 John K. Anderson GA (619) 455-2523 Leo Beltracchi RNC/RES (301) 492-4563 Paul R. Kasten ORNL

'(615) 574-6093 Jerry J. Swift **

NRC/NRR/DREP 492-7569 E. Chellith**

NRC/RES 492-8048 Fred A. Silaby GA (619) 455-4320-

- Joe F. Petrozelli*

Bechtel

'(415)'769-3129 4

Dave Dilling*

Bechtel

.(415) 769-3129 Robert Kirkwood*

NRC/RES William C. Craig*

SWEC John H. Flack NRC/RES (301) 443-7767 I

. May 27, 1987 only

• • May 28, 1987 only

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Enlosurel2-L

'l AGENDA MAY 27, 1987 i

NRC/ DOE MEETING ON MHTGR PSID CHAPTERS 6, 9, 10, 11, 12 and 13__

8:30 am Opening Remarks - NRC, DOE, and Others 8:45 - 9:15 am Overview - DOE, and Others 9:15 - 11:00 am Nuclear Island Building and' Structures Chapter 6 - BNI-Layout Overview

" Safety-Related" Structures-Functions and Requirements

-Design. Description

- Seismic

- Post-Break Venting Additional agenda items or discussions will include the following responses (1) res and clarifications:

12, 1987, (2)ponse to General Comment 11 in Morris to Gavigan letter of May how the blow-out panels work and their desi bases, (3) is the duct simply concrete or metal lined (PSID page 5.3-3)gn.

7 (4) is " controlled venting" of the reactor building considered a " safety-grade" operation? (5) explain the basis for assuming a 30'sq. in. leak.in the reactor coolant pressure boundry (PSID page 6.1-6) in. setting the vent path and (6) the role of the reactor cavity system in the attenuation factors given on PSID page 4.2.5 and discussed in Response 4-8.-

11:00 - 12:00 Coments by NRC and Discussion 1:00 - 4:00 pm Nuclear Island Service Systems (Chapter 9).

• Helium Purification System (HPS)

.GA

• Reactor Service Subsystem GA

• Fuel Handling and Storage System (FHSS)'

GA

• Site Fuel Handling

.DNI

• Reactor Plant Cooling Water Subsystem (RPCWS)

BNIL' Spent Fuel Cooling System-

'BNI

• NI Fire Protection-BNI

• NI HVAC BNI The NRC review goals for this and the two'following_ chapters are tolassureL.

that (1) all auxiliary and service equipment and. procedures that are related to Part 100 guidelines, whether on or off the-nuclear island, have been' identified, (2) the design is adequate to meet Appendix.1.and Part 20 requirements, Land (3) items identified as non-safety related do'not impair.by their failuref

'or by improper operation any equipment or procedure'important to safety.

In:

order to expedite the presentations and discussion _of.these items, the l emphasis should be on similarities, improvements or differences of equipment and procedures found acceptable for Fort St. Vrain or LWRs; as appropriate.

Particular concerns with the agenda items are (1).HPS - lack of principal design criteria, consequences of failures, roles of the LNS and chilled water systems in protecting against failure, (2) Reactor.' Service Subsystems - while these systems are important to plant operations the NRC does not pl.an to review and comment on the mechanical design.of any of these system # at the conceptual design stage unless it is judged they significantly impact plant-safety.

(3) Fuel Handling and Storage - assurance of prevention of criticality when control rods are removed during(refueling, correctness of orientation and location of-freshly installed fuel, 4) RPCWS - failures, leading to reactor water ingress or in safety related systems e.g., neutron control assemblies (PSID page 9._1 -85), (5) NI Fire Protection --justification for departure from Appendix R, protection to vital electrical equipment, assurance of operator access to remote shutdown area, hazards.from fire off the nuclear island,

' identification of " backup fire protection' services" (PSID 9.1-119), consequence of failures in f. ire. protection systems (including seismic) on meeting 10CFR100

. guidelines, interaction of. control room fire on plant shutdown and control..

description of the carbon dioxide subsystem, test and inspection intervals, independence'of fire pump diesel and support systems from plant ac power supply, (6) HVAC - potential as a release pathway from reactor. cavity during a LBEs, and (7) Other - safety classification of lighting, communi-cations, and instrument air systems.

4:00 - 5:00 pm Comments by NRC and Discussion

6 MAY 28, 1987 CONTINUATION OF NRC/ DOE HTGR MEETING, PSID CHAPTERS 10, 11,~12, 13 8:30 - 9:30 am Steam and Energy Conversion Systems (Chapter 10)

I main steam end feedwater systems. including l

" W ety-related" NSSS isolation valves 3t r tup / Shutdown *

• Steam Dump-Additional-agenda items or discussions should include: (1) possible.need for radiation monitors on steam vents and drains, (2) consequences to equipment on nuclear island.from turbine missiles, (3) design criteria,-functions and safety classification of the-service water system, (4) safety classification of the steam and water dump system.

9:30

-10:00 am Comments by NRC and Discussion 1

10:00 ~- 11:30 am Operctional Radionuclide Control (Chapter 11)

• Radionuclide Design Criteria *

• Rudwaste Systems

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• A00s*

Plant Dose Assessmin2 i

i Presentation and discussions:

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For the nuclides given in Table 11.6-1 plus the additional nuclides Te, Ru, Br$umn,headingsinasingletableorarrangementofsimilartables.*Ba, C-14 data and info]

Pu co

' Core Equilibrium - (T.11.1-1) l If this is different from total-inventory, then add a column " total inventory" and explain the difference.

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Core Design Inventory - (T.11.1.2) j Core Maximum Expected Inventory (T.11.1-3)

Plate Out The calculated amount-plated out.

-Plateout available for liftoff.- (T.11.6.1)

.If_this; column is the calculated goal then add also a column'of actually, anticipated plateout.

-Circulating Activity

-(T.11.6;1) q If this is the calculated goal then. add.also a columnLof' actual anticipated; circulating' activity.

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P For each column, explain how it was derived and what assumptions were made in the derivation.

In particular, where applicable, explain what role fuel contamination and failed particles play in the derivation.

Discuss the nuclides considered important from the standpoint of comprehensiveryss.

If-you believe any of the nuclides listed above are not important to release estimates and wish to drop them from the table, an explanation should be

- provided. Also be prepared to discuss if any other nuclides not listed on the table could be important.

In addition, clarification should be provided for:

(1) the significance of l

A00s and DBEs associated with Radioactive Waste and Fuel Handling and Storage Systems, (2) are any fission products retained by failed particles?' (if so

. give reference and quantify), (3) about half the helium is stated as not circulating-where is it? how does this. impact plateout?

(4) how is Fort St. Vrain experience being used to determine plateout activity and to estimate trituim release? and (5) how is the ALARA principal being applied?'

11:30 - 12:00 pm Comments by NRC and Discussion j

12:00 - '1:00 pm LUNCH j

1 1:00 - 1:30 pm.

Occupational Radiation Protection (Chapter 12)

Include (1) a description of shielding requirements for the helium purification system, (2) discussion of potential differences in the total exposure

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estimate from Fort St. Vrain, and (3) will R.G. 1.26 be met for radwaste systems.

1 1:30 - 2:30 pm Conduct of Operations (Chapter 13)

Plant Security Additional agenda items or discussions should include:

(1) review of position on emergency preparedness, (2) administrative controls including potential for and consequences of their failure.

2:30 - 4:0D pm Unresolved items Any or all of the subjects listed under Action Item-G-12 in the Morris to Gavigan letter of May 12, 1987 could be discussed at this time, 4:00 pm Adjourn I

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i Action Items and Clarifications i

NRC/ DOE Meeting, May 27 and 28, 1987 on MHTGR-PSID Chapters 6, 9, 10, 11, 12, and 13

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GENERAL COMMENT

G-13 DOE often uses the phase "will. meet the intent of" in committing the MHTGR design to.Various NRC licensing guidelines and regulations rather

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tnan making a full commitment. This is an acceptable and usually-1 desirable approach..~ DOE is requested to comment that its use of this phase is appropriate for one or more of the following purposes.and/or to suggest alternate.or additional language.

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.1 (1) The guideline or regulation was developed for light water reactors

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and contains some material irrelevant, inappropriate or incorrect for gas-cooled reactors, although gas-cooled reactors should' meet its objective in principle.

(2) Technical development of the guideline or regulation is. underway,.

contemplated, or likelyland full commitment to it at this state of.the design and licensing process would be premature,-unnecessary, and not in the best interests of safety.

(3) DOE plans to propose for a given item both a design. solution and relevant criteria that would be clearly a safety improvement over current j

practice.

(4) DOE plans to propose a relaxation in a current guideline or-J regulation for a given design item that can be shown by experiment, test, design, or analysis, or'an appropriate combination of these,.to maintain:

the level of safety at, or increase it from,' current standards. Where-DOEplansLsuchaproposalitsfulll justification.and;commitmenttoa-1 suitable plan for its support will'be required.

H We remark that in'no case would'the NRC permit.the use of "will meet the-intent of" to approve _the.adoptation of a guideline or reaulation.that o

would not maintain or increase the~1evel'of safety.

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i SPECIFIC C0tNENTS

'I 6-1 The intent of GDC 38, " containment heat removal" should be applied to the combination of the RCCS and those features of the reactor building that j

serve to assure that the containment function provided by the fuel particle coating is met. DOE should summarize how the intent of this GDC is being met b/ the MHTGR.

6-2 DOE should assess the potentials and probabilities for partial and full RCCS failures from such low probability events as an earthquake substantially exceeding the SSE, pressure, temperature, and heat loads greater than currently postulated, loss of essential d.c. power,

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significant delays in steam and feedwater isolation, intrusive fires, and l

the failure of sump pumps to remove a water intrusion in the cavity that occurs simultaneously with a need for the RCCS. The purpose of this assessment is to establish the margins to RCCS failure and the probability that the " earth-heatup" cooling mode would be needed.

In responding to NRC Comment G-11c, DOE should explicitly consider the effects of cavity insulation and the insulation on the RCCS panels in describing its position of the utilization of " earth-heatup" capabilities.

6-3 DOE should describe the potential for and the consequences of failure of the movable louvers in the reactor building to open in the event of a high i

energy line break.

6-4 DOE should document in the PSID, material presented-in viewgraphs and in-discussions pertaining to the reactor building performance. This docu-mentation should include the seismic analysis, response to pressurization, ventilation flow paths, and the credits to be taken for fission product attenuation.

E 6-5 DOE should discuss the conformance of the MHTGR nuclear' island with the intent of 10CFR50.49, " Environmental Qualification of Electrical i

i Equipment Important to Safety for Nuclear Power Plants." While we anticipate the discussion will emphasize " safety-related" equipment, the discussion should also include any non-safety related equipment for which credit may be taken in the PSID and the PRA. The discussion should J

specifically include the degradation of equipment as could be caused by

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high energy 1_ine breaks in the reactor building.

9-1 DOE should document in the PSID, viewgraph material-and discussions pertaining.to fuel handling and storage, particularly material summarizing j

the features that assure subcriticality during refueling and' features to-assure proper element replacement.-

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9-2 DOE should estimate the increase in dose for accident conditions that'

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might be caused by-(1) the misplacement of a fuel element with a' reflector element (that would cause flow blockage fo. the entire fuel' column) and (2) the misplancement of the the highest fissile content fuel' block in the

-highest flux _ location. DOE should discuss the potential' advantages of using a dissimilar arrangement of dowel pins and sockets for fuel--and reflector blocks to positively prevent thisLoccurrence.

9-3 DOE should document the basis for assuring suberiticality inithe spent ~

fuel storage wells.. Will subcritical multiplication experiments.be necessary to confirm calculations for storage of_'the-most reactive fuel elements (including fresh fuel in'the spent _ fuel storage wells)?-

9-4 DOE should identify (with comments as necessary) the' Regulatory Guides-

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and SRP sections that may apply to'the fuel' handling machine..

9-5 -DOE should describe the Seismic. Design requirement of.ith'e Spent-Fuel Cooling Subsystem.

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. i 9-6 While the Fort St. Vrain Reactor Plant Cooling Water subsystem.(RPCWS) is

" safety related," DOE stated that the reliability of the RPCWS for the-MHTGR will "be provided as needed."

Indicate where a reliable RPCWS may be needed (such as to protect or to maintain the availability of the HPS)

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and how the system will be designed to provide this rel'iability-(e.g.,

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type of electrical power supply).

9-7 For a " worst case" Helium Purification System (HPS) failure, DOE should 5

documenttheaccidentdosecons$quencesgivenattheEABandestimatethe i

worst case doses that could occur with respect to meeting the limits of l

10CFR20 and 10CFR50, Appendix 1.

Is there any case of HPS failure that-J could cause habitability problems related.to operator access or occupancy of'the remote shutdown area? Describe the various possible failure modes-of the HPS including the effects of failures of the RPCWS and the liquid 1

nitrogen system.

Include a description of the similarities and differences between failure modes of the HPS for the MHTGR and the Maximum:

j Credible Accident for the HPS analyzed for Fort St. Vrain.

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9-8 DOE stated that the Nuclear, Island' Fire Protection. System.will meet the.

intent of GDC 3, " Fire Protection," but that Appendix R'is intended for 1.WRs and not liTGRs. DOE further stated that components will.have

" reliability as required" and that the design will'be updated.and completed in preliminary and final design phases..To aid'our review in-this area DOE should describe how the MHTGR Safety functions can be j

performed given a nuclear island fire without a safety grade fire protection system in order to help support its statement that the fire i

protection system, is not relied upon to meet 10CFR100 requirements.

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9-10 The PSID stated that-the HVAC system does not perform any 10CFR100-related radionuclide control functions but that the system will have the appropriate q

reliability to meet other regulatory criteria. To aid our review in~this area please describe:

(1) Why HVAC failure or improper operation does not' result in a part 100 release, (2) How the design relates to meeting Appendix.I and Part 20 requirements, (3) Why failure or improper operation

'of-the HVAC system does not impair any equipment or procedure important to safety, and (4) the requirements for filtration systems.

9-11 The following items concerning the HVAC design should be documented.in

~the PSID: (1) the means to bring the chilled water to below ambient temperatures, (2) a description of the closed circuit; cooling unit for.

normal cooling of the steam generator cavity, and (3) the' reason that the Rad Waste building has an HVAC system separate from the main HVAC system.

H 9-12 DOE should document in that PSID that the helium storage and transfer q

system, the liquid-nitrogen system, t_he decontamination system and. the -

hot service facility are not relied 'upon to meet Part 100 requirements, do not initiate or aggravate any LBE, and.are systems similar.to those.

j of Fort St. Vrain.

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9-13 DOE should document discussion supporting the safety classification'of the lighting, communications-and instrument andl service. air'. systems.as-

"not-safety related".

NRC accepts DOEs' commitment-to' continue'to review these systems'as the design progresses.

DOE should document that the security system, including exterior lighting. would be on'en uninterrupted 1

power supply system.

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4 I 9-14 DOE should demonstrate that failures of any. auxiliary system (s), as could j

be caused by a large earthquake, will not impair operator access to the remote shutdown area or other plant. locations where' actions pertaining'to Part 100 releases may have to be performed. The auxiliary systems of particular importance in this demonstration are seen as HPS, the fire protection systems, the HVAC, the lighting system, and the communication-system.

8

.10-1 DOE stated that radiation monitors will be installed on the air ejector j

and the condenser.

DOE will document-reasons why there is to be no-radiation monitoring for the steam vents and drains.

I 10-2 The service water subsystem supplies cooling water to'the reactor plant cooling water system, the spent fuel cooling system, the chilled water' system and through a separate supply train to the shutdown cooling' system. DOE should document this overall arrangement in a schematic diagram in the PSID.

1 10-3 DOE should document in the PSID material presented describing the feedwater and steam isolation valves. DOE will address the, common mode failure potential of these valves at a later design stage..

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10-4 NRC will withhold judgment on the safety classification of the steam and water dump system, in.a manor consistent with Comment 5-29, until an.

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evaluation is made of the accident analyses.to be presented by DOE in the j

review of PSID Chapter 15. Our evaluation will also. consider compliance with GDC 10 " Reactor Design," and GDC 11. " Reactor Inherent Protection,'.'

and the potential for. consequences beyond;those analyzed in'the PSID.

11-1 The PSID, states that as the design progresses, appropriate'. evaluations.

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.. hile'this o

will be made'of release': sources outsi6-the reactor' building.

W planis acceptable in general, we request that'the~fo116 wing discussions i

be provided at this time with. respect.to failures in the radwaste system:

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• 1 (1) Could failure of any portion of the radwaste system and in particular the waste gas surge tank prevent operator access to the remote shutdown y

area or other plant locations for which access is required to releases below Part 100 guidelines?,

(2) Section 11.3.5.1.4 of the PSID suggests-y that the radionuclide inventory in the waste gas. surge tanks'will be' j

l limited so that if the contents of,one such component are released, the' l

maximum individual offsite dose will not exceed dose limits specified in 10CFR20. Are the radioactive waste. treatment components going to be sized q

to achieve such dose limits, or is this to be accomplished by adminis--

trative controls?

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11-2 If the Reactor Building Ventilation System is to be shared.among all four j

modules (PSID Section 11.3.1), what provision is made to limit the'

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impact on the other 3 modules if there is'a contaminating event in one module?

11-3 DOE will describe how Fort St. Vrain experience has been used in the e

design of the radwaste system and in estimating the amount of tritium to be released.. How will the disposal of. tritiated water be controlled

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in the MHTGR?

11-4 DOE will document in the PSID, material presented'on'how the ^LARh principle is being applied to the MHTGR.

12-1 DOE should clarify if the occupational exposure dose assessment for ISI also includes. doses from inservice testing.

If the dose estimate does not, then doses resulting from testing should be provided in Table 12.4-1 of'the

-PSID.

12-2 DOE should document in the PSID_a description'of the"related

and. unrelated aspectsofFortSt.Vrainexperienceused-in' making}theMHTGRoccupational' dose assessment.

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13-1 DOE reiterated its position that the emergency plan for the MHTGR would not require for the public outside the. plant site boundary, rapid notification, sheltering, or evacuation. We cautioned that the current bases for the NRC's emergency planning requirements contain non-It was also mechanistic elements which could be difficult to remove.

noted that 10CFR50.47b(2) states that "the size of the Emergency Planning Zones also may be determined'on a case-by-case basis for gas cooled nuclear reactors..."

13-2 Of concern to tlhe staff in its review of Chapter 13, Section 13.2.1,

" Philosophy of Plant Operational-Control," is how the HTGR meets GDC 13,

" Instrumentation and Control," and GDC 19, " Control' Room." We request DOE to support and/or augment its discussions of GDCs given in Amendment (1) Describe the design process used to relate the general 1

1 as follows:

l goals and top-level criteria of the design into the man-machine interface-design requirements for:

(a) the co'ntrol room;.(b) the remote shutdown panel; and (c) the PPIS equipment room. For each of these man-machine

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Justify interfaces, identify the functions allocated to human operators.

any exceptions taken from-the requirements of GDC 19.

(2) Describe in q

detail the role of operators to mitigate failures in plant systems and to mitigate the consequences of accidents.

In describing the role of.the operator, identify the features of the man-machine interfaces:that respond to the requirements of GDC 13.

This letter 13-3 NRC will prepare a special letter to DOE on plant security.

Lwill contain identification of designers'.and owners' responsibilities-at' the current stage of review, what portions of discussions and.submittals-should be restricted from public~ disclosure, and certain matters-of a technical nature.,A limited meeting may be held with DOE. prior lto the.

issuance'of the letter.

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13-4 Technical Specifications and other administrative controls will in l

i At general not be described or reviewed at the PSID stage of review.

l later review stages submittals should be in a format to address the applicability of the SRP.

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