ML20237G412
| ML20237G412 | |
| Person / Time | |
|---|---|
| Issue date: | 08/28/1987 |
| From: | Morris B NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | Gavigan F ENERGY, DEPT. OF |
| References | |
| NUDOCS 8709020238 | |
| Download: ML20237G412 (14) | |
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[f UNITED STATES
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p, NUCLEAR REGULATORY COMMISSION 3
?l WASHINGTON, D. C 20555
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August 28, 1987 l
Mr. Francis X. Gavigan, Director Office of Advanced Reactor Programs Office of Nuclear Energy U. S. Department of Energy Washington, DC 20545 p
Dear Mr. Gavigan:
On August 12, 1987, members of the NRC staff and its contractor (BNL) met with members of DOE and its contractors to review Chapters 6 and 9-13 of the SAFR j
Preliminary Safety Information Document (PSID) - Project 673. The agenda and list of attendees are given in Enclosures 1 and 2, respectively. The meeting consisted principally of presentations by DOE's primary contractor (Rockwell International - RI) for this project on containment design described in Chapter 6 of the PSID and the various systems described in Chapters 9-13 of the P5ID. The NRC staff comments on these chapters were discussed.
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i The action items, including requests for additional information and/or clarifi-cation, that resulted from the meeting are given in Enclosure 3.
A set of preliminary questions were submitted informally prior to the August 12, 1967 meeting. These provided a basis for structuring the meeting and were responded i
I' to in some detail at the meeting. There have been some minor changes to some of these questions to clarify and, in some cases ~, to expand on the question.
Such questions are marked with an asterisk. Several additional questions resulted from discussions at the meeting.
These are:also included in Enclosure 3.
f Your response to the enclosed questions is requested by September 25, 1987 in order to maintain our review schedule.
Sincerely, l
kcusBill M. Morris, Dire D ls s Division of Regulatory Applications Office of Nuclear Regulatory Research
Enclosures:
1.
Meeting Agenda 2.
List of Attendees 3.
Questions
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Au' gust 28, 1987 c:
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' DISTRIBUTION
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RES Circ Chron e
ARGIB R/F E~. Beckjord T.LSpeis B. Morris Z. Rosztoczy T. Xing J. N. Wilson C.' Allen' R. Landry M. Dey P. Williams'.
J. Flack.
R. Baer N. Anderson F. Cherny S. Shaukat R. Johnson D. Thatcher J. Hulman J. Glynn j
L. Soffer J. Read D. Cleary G..Arndt.
R. Kirkwood R.'Erickson B. Mendleschn 1
E.'Chelliah.
1 t
L. Beltracchi
!F. Congel L.'Cunningham J. Minns D. Matthews F. Kantor E. Podolak M. Spangler G. VanTuyle, BNL M. El-Zeftawy, ACRS/H-1026 j
PDR - Project 673 w/o enclosures 4 Project;Fj le g.673;;,( Centra l;F i les ) gma t-4 j
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ENCLOSURE 1 AGENDA FOR AUGUST 12 NRC REVIEW r,.
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Presentation Scheduled 1
P510 pg subhet Tise Tiw _
Introductlen
(!ncludes 9:00a.m.
respones to presubaltted questions) 6.2 Cantalinant w
9:10 a.m.
Primary secondary 10 9:55 a.m.
9.1 Fuel Storage and h aditne 20 10:10 a.m.
9.2.2 Aurilla cooling Water System 10 10:40 a.m.
(
is an secondary con inment) 10.55 a.m.
BREAK 10 11.05 a.m.
9.4.1 pl WAC tystaa 10 11:20 a.m.
9.4.2 Control Sullding WAC 10 11:35a.a.
9.7 kbitabilitySystem 9.5.1 Prine / Sodlun Processing 10 11:50 a.m.
12:00 acon tiscH 9,6.1 Ftrv Proteuttw Snt=4 Sullding,and(NI, Control Conventional 15 1:00 p.m.
Security) 20 1:20 p.m.
Sodium 10 1 50 P A 9.G.2 Cennunleation Systas 15 2*05 p.m.
1.6.3 overhead Heavy toad h adllt.9 System (Descriptions only - no PAA discussion at this time) 5 2:25 p.m.
l 11.1 1.lquid hsta t%tnagement systens 5
2:25 p.m.
j 11.3 Solid Waste Mugement Systens 15 2:35 p.m.
11.2 cueou: Waste Management Systen (Emphasis on storage) 11.4 Process and 2ffluut Rat 01091 5
2:55 p.m.
cal Monitoring and 5mp1Lne j
Systmo in 2:05 m m.
17.1 M W W rewyine 13.3 Safeguards & 59eurity Planning
_30 3:20 p.m.
(use utility presentation) gg 4:00 p.m.
l DISCUSSION f
4:30 p.t.
l ADJOURN 8822A/ reg 1
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ENCLOSURE 2 MEETING ATTENDEES CHAPTERS 6, 9-13 0F SAFR PSID AUGUST 12, 1987 NAME AFFILIATION TELEPHONE CeMis Allen NRC-RES (FTS)492-8302 Zoltan R. Rosztoczy NRC-RES 443-7566 Moni Dey.
NRC-RES 492-8106 Dean R. Pedersen Argonne
-(312)97c 3335 George Sherwood DOE (FTS)233-4162 Jerry Wilson NRC-RES 492-4727 Greg VanTuyle BNL (FTS)666-7960 Tom King NRC-RES (301)492-7014 John H. Flack NRC-RES (301)443-7767 L. N. Rib LNR Associates (301)983-1032 B. T. Mendelsohn NRC-NRR/DRIS (FTS)492-9671 J. Ross Humphreys DOE 233-3588 John Hagelston RI (818)700-3282 Richard E. Johnson NRC-RES 492-8129 R. (Bob) DeRusseau Bechtel (415)768-8555 Charles E. Jones RI (818)700-3282 John L. Minns NRR-RAB-PRPB 492-4086 1
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ENCLOSURE 3-NRC Questions on SADR PSID Chapters-6, 9 6.1 The: secondary containment is rot currently designed for long term retention rf releases in that the design leakage rate is 100% vol/ day.
Tables-15,2-10 and 15.2-11 in the PSID indicate that for'the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 24' hour doses associated with RI's proposed site suitability source term the secondary containment reduces the offsite dor,e by:
approximately a factor of ten. For the 30 day dose it appears the secondary containment has no effect.
To b'etter understand the' role of.
the secondary containment and the assumptions made in the dose analysis please describe:
a)
What credit,-if any, is taken from fission product plateout, filtration, chemical reaction, etc. in the secondary containment prior to release to the atmosphere (based upon the 30 day dose values in Tables'15.2-10 and 15.2-11 it appears no credit is taken)?
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If no credit-is taken in (a) above, how would the dose rates in Tables 15.2-10 and 15.2-11 char.ge if a realistic assessment is made?-
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If credit is taken in (a) above, what is the basis for the plateout, filtration, etc. values chosen and how will it be demonstrated that they can be met? Also, what would be the dose rates in Tables 15.2-10 and 15.2-11 if no credit were l
taken?
l Has RI considered options to reduce the release rates from the secondary containment through the use of liners, seals, welds, etc.
to provide additional margin?
6.2 How large a gap between the reactor vessel and the guard / containment l
vessel is necessary for your planned in-service inspection program?
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- 6.3 Recogiiizing the geometrical complexity of the closure-to.the SAFR vessel,. will DOE meet the ASME Code Section XI ISI requirements? If so, describe them and the proposed means of meeting them; if no (or for any which will not be met), explain why not.
- 6.4 At the meeting on August 12, 1987 the statement was made (viewgraph 870-14-1662A) that there is a " substantial in-vessel core debris coolability (i.e., potentially 100%) fnr hypothetical events." Please-provide the basis for that statement.
- 6.5 Please provide additional information (figures and description) of the boundary between the-secondary containment and the fuel handling cell.
l9.1 The cold trap device, described in Section 9.5.2.2, utilizes helium as 1
a coolant. Can this helium be released into the cold pool or reactor inlet plenum? If so, what is the maximum amount of helium, and the corresponding reactivity worth, that can be introduced into the reactor?
9.2 How does the operator recognize loss of cold trap cooling, and what actions must be taken and in what time frame?
9.3 In the case of a large sodium fire, how will the reactor be shut down?
Is operator. action required in some cases?
- 9.4 Describe the contingency pian for the situation where the fuel transfer cell handling machirie becomes inoperative during transfer and discuss how the proposed design will minimize such failures.
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- 9.5 Please provide additional information on the flow paths of the air for cooling in the interim fuel storage cell. Also, provide an assessment of the effects from blockage of these flow paths.
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- 9.6 Describe the provisions to monitor temperatures in the interim fuel storage facility, including temperatures of the stored fuel.
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- 9. 7 Provide an assessment of the capability of the SAFR design to with-stand a conventional fire and still perform its safety functions.
Include considerations of electrical disturbances such as " hot shorts" in your assessment.
- 9.8 Provide an assessment of the effects from failure of the fans and louvers in the steam generator building, i
- 9.9 We have the following concerns with regard to the non-sodium fire protection system described in Section 9.6.1.1 of the PSID:
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(a) the water source is not classified as Seismic Category I nor is there a backup supply, (b) there does not appear to be an automatic suppression system for the IE battery rooms, Until we have a much better understanding of the response to the SAFR plant to fires (see question 9.7 above) we could not approve such a fire protection system design.
Your response to this comment may be combined with that for question 9.7 above as appropriate.
i 10.1 The SAFR Steam and Power Conversion System looks very much li'ke a light water reactor (LWR) rystem, modified for the high pressures and temperatures present in SAFR. A large fraction of.the transients in an LWR system originate in the B0P, so presumably the same will be true in SAFR. Do the duty cycle, the PRA, and the plant availability assumptions rcflect this?
11.1 On Page 11.3-3 (second paragraph), it is stated that "approximately one or two control rods are replaced each year." Why is it necessary to replace rods so often (high burnup of the boron carbide)? Is the control rod worth droping significantly during a one year period? i
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- 12.1 Do-the dose rates given in Figures 12.1-2 through 12.1-11 for areas near the fuel handling cell air exhaust stack and the DRACS and RACS air exhaust stacks include contributions from radioactive materials in the exhaust air itself? How significant would such contributions be with and without consideration of failed fuel. pins?
13.1 The Security Plan
- was submitted under provisions of 10CFR2.790(d).
That regulation applies to pnysical protection information not other-wise designated as Safeguards Information or classified as National Security Information or Restricted Data.
On-the other hand, documents containing lists.or locations of equipment explicitly identified in the documents as vital equipment.or vital areas, in the context of 10CFR73.55, as 'well as details of the reactor physical protection sys -
tem, may need to be identified as Safeguards Information and protected in accordance with 10CFR50.34(e) and 10CFR73.21. These requirements should also be taken into account in responding to the following
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questions on safeguards i.nd security.
a) The Commission's Severe Accident Policy Statement states:
"The Comission also recognizes the importance of such potential contributors to severe accident risk as human performance and sabo-tage. The issues of both insider and outsider sabotage threats will be carefully analyzed and, to the extent practicable, will be emphasit.ed in the design ~and in the operating procedures developed for new plants."
Also, new applicants are to describe how they will address certain generic safety issues, including A-29, " Nuclear Power Plant Design i
for the Reduction of Vulnerability to Sabotage". The discussion of A-29 in the PSID's Appendix B, which deals with generic safety l
issues, describes SAFR design features that would make these plants more inherently safe against insider and outsider sabotage. The l
- M. E. Remley, et al., "SAFR Safeguards and Security Plan and Assessment,"
Rockwell International Report 140SS000001, Rev. A (December 20, 1985).,
PSID's Appendix C, on the PRA, makes no' attempt to quantify these risks using even conditional probabilities; its section 6.6, entitled " Sabotage", instead provides an overview of the security design concept..That section should,-as a minimum, describe how RI intends to quantify such risks.
Eb) The Task Statement for Generic Safety Issue A-29 states:
... a need to investigate measures to balance safeguards interests with plant safety interests. This.is an ongoing concarn in light of experience gained and questions raised from various sources regarding the potential interference of security measures with safe operations, especially in reactor upset conditions."
Page 64 of the Security Plan states:
" Normal personnel access [into the Nuclear Island (NI) building]
would be at the building personnel portal, where positive personnel-recognition is required."
Could " positive personnel recognition" devices cause delays in entering the Nuclear Island? It is possible that longer time constants before intervention is required in response to upset conditions may mitigate potential concerns about assuring timely access in upset conditions, but it needs to be addressed in the PSID.
c) As some equipment within the nuclear island may be vital, consideration of the needs of vital barriers during preliminary building-design could be beneficial. Consideration should be given to the definition of a vital area barrier in 10CFR73.2(f)(2) and to the regulatory position on openings in physical barriers in f
Regulatory Guide 5.65, which could impact on design of some ducts and penetrations.
In particular, air intake ducts for the reactor air cooling (RACS) may need to be designed to prevent their use for l.
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' unauthorized access to the containment vessel (CV) surrounding the l
reactor vessel (RV). Would barriers such as depicted.on Regulatory Guide 5.65 page 11 have an affect on the loss coefficients presented in Section 6.3 of the PSID? Similar barriers could be J
needed for ducting to the RACS inlet ducts from the HVAC in the 1E I
Vital Instrumentation AC Room if that ducting was man-sized (i.e.,
96 square inches with.at least one dimension equal to or greater than 6 inches).
d) PSID Table 6.4-1 states for RACS that:
" Personnel may enter all passages for direct VTM-1 visual inspection."
Is this consistent with page 41 of the Security Plan that says the only openings to RACS are through the RACS vents? What security measures would be used to protect the CV from the inspector; from potential saboteurs?
e) Will security measures be provided on reactor containment building (RCB) roof hatches (which may be periodically opened for equipment removal), HVAC vents, and maintenance access ports?
f) The PSID notes that the guard piping over sodium piping in the RCB
-is there to protect against sodium fires in containment. What attention has been given to protection against deliberately caused sodium fires?
g) PSID Table 5.6-5, Component Reliability Data, lists RACS unavailability and failure rates as very small but finite values.
Have unavailability / failure mechanisms been identified, and could those mechanisms be exploited by insiders?
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h) A security advantage is claimed becwse only the nuclear island is t
in the protected area, not the balance-of-plant (BOP) equipment:
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"This approach severely _ limits the number of personnel permitted in each zone and provides a major deterrent to the insider internal threat." [page 12, Security Plan]
But page.12 also states that BOP personnel may be granted unescorted access to the NI protected area if "specifically authorized." Page 60 says they can have unescorted access if "their work functions require routine access." How will " require routine access" be defined to significantly limit the number of B0P personnel who justify being "specifically authorized?"
,y i). There may have been some design changes not incorporated in Revision A of the Security Plan. The PSID (Revision 11) shows two, not four, RACS exhaust stacks.
Does this affect the feasibility of blockage? Also, there is no steel l ner on the RCB concrete i
structure according to page 3.8-2 of th'e PSID, while page 43 of the<
j Security Plan says "the containment building is a 3-ft-thick concrete structure with a carbon steel lining." Please clarify these differences
- 1) What evidence is there to. support the contention that an explosive attack from offsite on the RACS and DRACS exhaust towers would not block the requisite i ow?
k) We disagree with the position stated on page 34 of the Security Plan that only the RV and not the CV would be considered vital equipment. We. note that an explosive charge on the CV could also breach the RV. This point was also raised in a letter from T.
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Speis (NRC) to Francis X. Gavigan (DOE), dated July 28, 1986, which stated that we believe containment should be identified as a system requiring protection.
Your response should address that comment as well, if it has not already been answered.
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- 1) Page 36 of the Security Plan states that the Central Alarm Station-(CAS) is not considered a vital area.
10CFR73.55(e)(1) requires CAS to be designated a vital area.
m) -10 CFR 73.55(c)(2) requires physical barriers at the perimeter of 4
the protected area to be separated from vital area barriers.
Plan views of the plant appear to show NI and other buildings to have
. walls of buildings, that could be vital barriers, to be part of thi inner NI protected area' barrier. This would also appear to conflict with.the requirement for an isolation zone on both sides of the protected area fence. Also, the drawings do not match the
' description of the perimeter in 2.6.3.1 of the Security Plan. Does the inner fence, its intrusion detection system, an interior L
iisolation zone and an interior patrol road surround the entire y
protected area as stated in 2.6.3.17 Would patrols of the perimeter be encumbered by the plant layout shown in the drawings?
n) As noted in our July 28, 1986 letter from Themis P. Speis to Francis X. Gavigan, the physical security system will.have to
-include additional features for protection against the insider threat, including measures to' assure trustworthiness of persons with. unescorted access, and search of persons entering the
. protected area, i
c)
It'is premature to determine the armed response force size and
. total number of security officers that will be required.
For planning purposes we recommend using the regulatory nominal of 10 armed responders, since NUREG -0907 requires knowledge of site specific information and procedures not available at this stage, p) The 1.0 fc on page 62 of the Security Plan should be 0.2 fe, as re-quired by 10CFR73.55.
It'is correct on page 56.
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. q) Applicable rules.and regulations include 10 CFR Part 73, %% 73.20,'
'73.45, and 73.46, for protection of strategic special nuclear material from theft.
17,1 Please respond to the following questions regarding the quality
.M assurance (QA) programs used in producing your PSID. Although there is l
rio Chapter 17 in the SAFR Conceptual Design PSID, we have labeled these j
ovestions as 17.X consistent with the quality assurance chapter in a
' a) The enclosure 149QPP000001' entitled "SAFR Quality Assurance Plant to the 5/22/87 letter from F. Gavigan (DOE) to B. Morris (NRC) g
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appears to have been first issued in 4/85.
Please identify any work on the PSID that was performed prior to that date and describe the QA programs that were applied to the work done during that 3
- time, b) Section B of the 4/20/87 letter from R. Lancet (Rockwell International)'to F. Gavigan (DOE) indicates that the SAFR QA program at RI addresses the pre-docketing areas of review identified in Chapter 17.1 of NUREG-0800 (except for procurement document control and several other similar areas) and even goes beyond the pre-docketing areas of review. With regard to this aspect of your program we need to know:
(1) to what extent did your associated contractors (Bechtel National and Combustion Engineering) follow the QA program applied by R1, aii8 (2)'
does the QA program referred to above include application of all of the relevant SRP acceptance criteria (1,2,3,5,6, 16,17 and 18) by both RI and its associate contractors,
.Bechtel National and Combustion Engineering?
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c) Section D of the 4/20/87 letter from R. Lancet to F. Gavigan describes the audit activities at RI.
In paragraph 2 you indicate that the QA procedures rather than detailed technical reviews of designs and analyses were performed. While paragraph 1 indicates that DOE does the technical review, this seems to indicate that RI is not meeting the requirements of RG 1.28 Revision 3, 10 CFR Part 50 Appendix B and NQA-1 which requ' es audits to determine l
l compliance with and effectiveness of the QA program. Does DOE consider this arrangement acceptable, and why? Also, please describe the scope of the audit activities performed by your associate contractors'Bechtel National and Combustien Engineering.
d) Paragraph 3 of Section C of the 4/20/87 letter from R. Lancet to F.
Gavigan indicates that EMP 3-63 provides control of computer codes and calculations at RI. To what extent does this apply to work done by your associate contractors Bechtel National and Combustion Engineering? If not, how do those organizations control computer codes? Also, since EMP 3-63 was released in 5/84, please identify work performed prior to that date and describe the QA controls on computer codes and calculations applied to that work.
e) Please indicated which portion of Chapter 17 of the LWR-SRP will apply to SAFR during final design, construction and operation, provide justification for any deviations from SRP requirements..
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