ML20235C912
| ML20235C912 | |
| Person / Time | |
|---|---|
| Site: | 05000000, 05000447 |
| Issue date: | 03/20/1974 |
| From: | Stolz J US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Hinds J GENERAL ELECTRIC CO. |
| Shared Package | |
| ML20234E460 | List:
|
| References | |
| FOIA-87-40 NUDOCS 8707090577 | |
| Download: ML20235C912 (42) | |
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ff G--*, ATOMIC ENERGY COMMISSION F/ INFO ACES WASHINGTON, D.C. 20545 f
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MAR 2 01974 Docket No. STN-50-447 l RECEIVED
- gg,3r;R1 C0lWITTEE ON U.S. A.E.C.
REA010R SMEGJAhC3 General Electric Company MAR 2 5 Wh I ATTN: Mr. John A. Hinds, Manager PM 1 Safety and Licensing Igiggi 2g3g4 t ih6 175 Curtner Avenue i )
San Jose, California 95114 d Gentlemen:
In order that we may continue our revie,w of your GESSAR application I additional information on those matters set forth in the enclosure is needed. l To maintain our licensing review schedule, we will need a completely adequate response by May 9,1974. Please inform us within 7 days af ter receipt of this letter of your confirmation of the schedule or ,
the date you will be able to meet. If you cannot meet our specified j date or if your reply is not fully responsive to our requests, it is I highly likely that the overall schedule for completing the licensing !
review for this project will have to be extended. Since reassignment of the s taff's ef forto will require completion of the new assignment prior to returning to this project, the amount of extension will most likely be greater than the extent of delay in your response.
The questions in the enclosure have been grouped by sections that correspond to the relevant sections of GESSAR, and the question numbers continue consecutively from the December 12, 1973 letter.
Please contact us if you desire additional discussion or clarification j I
of the material requested.
E$ incerely, O ,ty , ,9,.g,
- 7 s p.t el e /
h Ebh.
d Jo 'n F. Stolz, Chief J ght Water Reactors Branch 2-1 Directorate of Licensing
Enclosure:
GESSAR Request for Additional Information ,
cc: (see next page)
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G707090577 G70623 PDR FDIA k, h &, ana r THOMASB7-40 PDR 4
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e i General Electric Company -
2-IGR 2 0 94 cc: Mr. W. Gilbert, Manager Safety and Standards General Electric Company 175 Curtner Avenue San Jose, California 95114 Mr. L. Gifford, Manager Regulatory Operations Unit General Electric Company
- 4720 Montgomery Lane Bethesda, Maryland 20014 9
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ENCLOSURE, REQUEST FOR ADDITIONAL INFORMATION GENERAL ELECTRIC COMPANY GESSAR PLANT l DOCKET NO. STN 50-447 i
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2.0 Site Characteristics 2.8 Your response to Question 2.6 implies that the safety design basis of the standard plant is such that the 30 day doses beyond the exclusion areas are significantly below the guidelines of 10 CFR Part 100. Resolve the discrepancy between that statement and the 30 days dose curves of Figures 15.1. 39-1 and 15.1. 39-2.
?.9 The response to Question 2.7 is not adequate. Discuss the inherent protection provided by the GESSAR design against aircraf t hazards.
The discussion should be independent of 'the location of the plant and should conclude with the characteristics (weight, velocity, etc.) of tije worst aircraf t and missile impacts that the plant can withstand.
On the basis of current information GESSAR is not acceptable at any site there the probability of a strike by any size aircraf t is greater than 10 7 per year.
'e .1-I 3.0- Design of Structures. Components. Equipment and Systems 3.140 Your response. to question 3.87 concerning the target response to missile impacts utilizes reference 17 which is ANSI NI77. This document has not yet been approved by'the Regulatory staff. Acceptable methods of predicting overall damage are given in Section E'.2 of Attachment II and referenced in question'3.98. Compare the procedures delineated in ANSI NI77 with those of Attachment II and justify the deviations in methods.
3.141 The- text of Section 3.7.1.2 (Amendment 8) is incomplete. This section should be revised to provide the response spectra which envelope the design response spectra shown in Figures 3.7-1 and 3.7-2, for all damping values to be used in the design. The spectrum period intervals at which the spectral values -
were calculated should also be provided. Additionally, the second paragraph of the section should be revised to show a comparsion of spectra instead of
~ comparing the OBE time history to the design SSE spectra. ,
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3.142 In Table 3.7.1, the asterisk signs pertaining to soil should be replaced by a 10% value for the lumped-mass soil spring approach. Furthermore, the footnote should reflect that shear strain dependent soil damping values, as defined in an attached figure (to be provided by the applicant) would be j used for finite element soil--structure interaction analysis. )
'l 3.143 Provision of seism'ic instrumentation, as described in Section 3.7.4 is j inadequate. It is our position that the following instrumentation should be 1
installed at appropriate locations in order to have an acceptable program )
i for seismic instrumentation:
A. One triaxial time-history accelerograph at each of the following locations:
- a. Free field,
- b. Containment foundation,
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- c. Containment structure or reactor building, at a higher elevation directly over the'one placed at th'e foundation.
B. One triaxial response-spectrum recorder or multi-element seismoscope at'each of the following locations:
- a. . Containment foundation - with capability of providing signals for immediate control room indication,
- b. Rea'ctor equipment or piping support,
- c. Category I equipment or piping support outside containment,
- d. Category I structure outside containment.
C. One triaxial peak accelerograph at each of the following locations:
- a. Reactor equipment, i a '
- b. Reactor piping,
- c. . Category I equipment or piping outside containment.
D. One triaxial seismic switch at the containment' foundation.
The location of instrumentation should be determined by dynamic analysis such that most pertinent data of seismic response can be obtained.
Response-spectrum recorders or multielement seismoscopes specified in Paragraph B may be replaced by time-history accelerographs except the one installed at the foundation level. The one at containment foundation i'
L should retain the capability of providing response spectrum data for immediate control room indication.
A plan for timely utilization of the data to be obtained from the installed seismic instrumentation should be developed.
3.144 The response to question 3.83 is unacceptable. We understand that the foundation medium properties data used in the soil-structures interaction analysis, and requested in the question, are currently being developed by and your consultant. Describe the procedures used to develop the requested
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. i 9 4 3 data and provide tentative schedule for submittal of the information.
3.14 5 Feview of your . response to Question 3.105 (provided in Section 3.7.1.5)
, indicates. that the embedment depths (measured to the bottom of the slah) for structures such as the reactor building, fuel building, radwaste build-ing'and auxiliary building etc., appear to be greater than the 30 foot embedment depth you plan to use in generating the GESSAR seisnde design env elope. It is the Stafis position that the embedment depth used in seismic design envelope generation shall be equal to or greater than the depth range listed in Section 3.7.1.5.
3.146 For. those portions of the containment and its floor liner plate that are submerged'in the suppression pool, your proposed design, (described in Section 3.8.2.4.4) using carbon' steel with corrosion allowance but with no protective coating, will be acceptable if the following are provided:
(a) Procedures and methods for maintaining clean, corrosion-retarding conditions in the suppression pool water, (b) An inservice surveillance and inspection program oriented towards detection of corrosion particularly with respect to pitting, and (c) Technical specification requirements for actions to be taken if exces-sive corrosion in detected.
3.147 It is a Regulatory Staff position that the drywell be structurally tested at 1.15 times the design pressure generated by a loss-of-coolant accident, and leak tested at about the drywell design pressure. Periodic low pressure ;
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- 1. leak tests should be conducted thereaf ter. {
3.148 With regards to post-LOCA flooding of the containment, the following informa-tion should be provided:
(a) An outline of the sequence of events that will be followed within the 1
design capability of the plant to ficod the containment and subsequently l' , ,
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remove the fuel.
(b) the stress limits that the containment can withstand considering flooding to the level of t.he operating floor without a simultaneous OBE, and (c) the feasibility of designing the containment to withstand flooding up to 7'-0" above the top of the core with an OBE, to within stress allowables of the ASME Section-III Code for Emergency Conditions.
If this is not feasible, the stress limits to which G. E. can reasonably design the containment,to withstand flooding to 7'-0" above the top of the core with an OBE without having to increase the thickness of the shell to an extent which will then require post-welding stress-relieving heat treatment.
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.. 4 3.149 Table 3.8.3 as revised does not contain the-impact due to jet impingement as required in NE3131.2 (Winter 1972 Addenda) and discussed in Regulatory l
Guide 1.57. Table 3.8.3 should be revised accordingly. Furthermore, consideration of f atigue should also be indicated.
3.150 Your response to Question 3.99 is incomplete. It states that figs. 3.8-12 through 3.8-17 will be changed to reflect the tornado and missile protection features. A schedule for the completion of these changes I should be provided. '
3.151 In your response to Q 3.94, the definition of Tt and Pt should be revised
" to indicate that the drywell will be tested at 1.15 times the full design pressure generated by.the loss-of-coolant accident.
3.152 Load combinations listed on page 3.8-22 for portions of the drywell that l
are constructed.from steel plates are not applicable. If the steel plates are intended to act compositely with the concrete between them, the entire section becomes equivalent to a reinforced concrete section and accordingly,
'the load combinations of Section 3.8.3.1.3.3 are more applicable whereby ,
l the plates may he treated as reinforcement.
i However, if the steel plates are intended to resist the loads alone without the concrete between them, they become equivalent to steel shell components and accordingly, the design criteria related to load combinations and allowable stresses of Subsection NE of the ASME Code Section III, Division I, as augmented, where applicable, by Regulatory Guide 1.57, are more applicable. Your intent to comply with these criteria should be indicated.
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3.153 Verify that the " minimum actual uniform elongation" and " minimum percent elongation" mentioned in Section 3.6.3.1.5.l(3)a under Design Local Strain for Type I Components are the elongations at the maximum stress on an engineering stress-strain curve, and not the elongation at the fracture stress.
3.154 Verify that the strain associated with "80% of the minimum calculated static ultimate restraint strength" discussed in Section 3.6.3.1.5.1(3)b under Design Steady State Load for Type I Components is less than 50% of the ultimate uniform strain for all materials which will be used for ,
Type 1 pipe whip restraints. Ultimate uniform strain is defined as the strain associated with the maximum stress on an engineering stress-strain curve, i.e. , the strain at which necking down of the specimen begins.
3.155 Supplement the response to Question 3.55, by providing a detailed description of the methods and procedures which will be used to postulate the jet impingement forcing function by utilizing the theory presented in the referenced Moody paper, including the magnitude variation with distance i
and time jet geometry such as trajectory and angle of dispersion, shape effects of the impinged object, and the loading distribution on the impinged surface.
In lieu of the above, the methods and procedures used in General Electric Report No. 22A2657 "Enrico Fermi 2 Pipe Whip Report" may j
l be referenced provided additional forcing functions to include the l
zone beyond five diameters of the ruptured pipe are desc rib ed. In l
1 addition, the use of 12-1/2 degree uniform angle approach for computing jet impingment loads within the 5 pipe diameter zone is acceptabic for l
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9-steam and steam / water mixed flow due to the existence of more severe Jet expansion under choked conditions. However, a 10 degree uniform angle is appropriate for sub-cooled water. The total jet impingment loads should not be less than the jet thrust force as computed .in Section A.of Report 22A2657. .
3.156 The response to Question 3.63 only discusses active pumps and valves and does not address the subject of both Questions 3.56 and 3.63, i.e.,
mechanical equipment. In Section 3.9.l.1.2 of GESSAR, provide a more detailed description of the seismic qualification program which will be implemented on all seismic Category I mechanical equipment and supports.
d Acceptable criteriaare outlined in Attachment III to Section 3 entitled
" Electrical and Mechanical Equipment Seismic Qualification Program."
3.157 Tne response to Question 3.64.1 is not adequate. In, addition to testing the recirculation piping system to the steady state operational limits, we require the system to be subjected to certain transient loads which result from valve closures, pump trips, etc. Acceptable criteria for such a program is outlined in Attachment I to Section 3 (previously supplied to you) entitled "Preoperational Piping Dynamics Effects Program."
3.158 The response to Question 3.64.3 is not adequate. As a minimum, we require a commitment to perform preoperational vibration tests on all ASME Class 1 and 2 piping systems consistent with the program outlined ;
in Attachment I to Section 3, "Preoperational Piping Dynamics Effects Program."
}159 The. responses to Questions 3.63 and 3.70 do not provide sufficient detail to enable the staff to properly evaluate the program outlined in Section
.5.2.1.7 of GESSAR. A program which is acceptable for plants currently undergoing review is contained in Appendix 3.9.B of the Allen's Creek Nuclear Generating Station PSAR Amendment 3 (Docket No., 50-466/467.) with the following additions:
(a) With respect to the analysis and/or testing of active pumps, the nozzle loads for the faulted condition shall be considered as the normal (design) condition pump nozzle loads and so specified to the pump manufacturer. The pump manufacturer should provide a d
pump suitable for normal operation with these end loads.
(b) With respect to the static shaf t deflection analysis of the motor rotor, which is discussed in Sect. 1.1 of Appx. 3.9.B. provide the basis for
, the 3g horizontal and lg vertical SSE accelerations acting simultaneously as input to the analysis. .
(c) With respect to the discussion of the qualification of safety-relief valves on page 3.9.B-5 of the Allen's Creek PSAR, include-a description of a test and/or analysis to demonstrate that a seismic event will not prevent the valve from opening, e.g., a test which applies a static load equivalent to the SSE to the top of the valve bonnet and simultaneously increasing the pressure until the valve mechanism actuates.
Revise Sections 3.9.2.4 and 5.2.1.7 of GESSAR to reflect this type of l detail for' all ASME Class 1, 2 & 3 active pumps and valves.
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. 12/5/73 Attachment TII to Section 3.
ELECTRICAL AND MECHANICAL EQUIPMENT SEISMIC QUALIFICATION PROGRAM l
I. . Seismic Test for Equipment Operability
- 1. A test program is required to confirm the functional operability l of all Seismic Category I electrical'and mechanical equipment and instrumentation during~and after an earthquake of magnitude up to and including the SSE. Analysis.without testing may be acceptable only if structural integrity alone can assure the design intended function. When a complete seismic testing is impracticable, a combination of test and analysis may be accept-able.
- 2. The characteristics of the required input motion should be specified by one of the following:
(a) response spectrum .
.(b) power spectral density function (c) time history Such characteristics, as derived from the structures or systems s . seismic analysis, should be representative of the input motion' i
'at'the equipment mounting locations. .
. 3. Equipment-should be tested in the operational condition. Oper- ,
' ability should'be verified during and after the testing.
- 4. The actual input motion should be characterized in the same manner as the required input motion, and the conservatism in amplitude and frequency content should be demonstrated. i l
S. Seismic excitation generally have a broad frequency content. )
-Rendom vibration input motion should be used. However, single frequency input, such as sine beats, may be applicable provided one of'the following conditions are met:
(a) The characteristics of the required input motion indicate that the motion is dominated by one frequency (i.e., by structural filtering effects).
(b)Theant'ibpatedresponseoftheequipmentisadequately represented by one mode.
(c) The input has sufficient intensity and duration to excite all modes to the required cagnitude, such that the testing response spectra will envelope the corresponding response spectra of the individual modes.
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- 6. The input motion should be applied to one vertical and one principal (or two orthogonal) horizontal axes simultaneously unless it can be demonstrated that the equipment response along the vertical direction is not sensitive to the vibratory l- motion along the horizontal direction, and vice versa. The time phasing of the inputs in the vertical and horizontal direc-tions must be such that a purely rectilinear resultant input is avoided. The acceptable alternative is to have vertical and I horizontal inputs in-phase, and then repeated with inputs 180 degrees out-of-phasa.. In addition, the test must be repeated I
with the equipment rotated 90 degrees horizontally.
- 7. The fixture design should meet the following requirements:
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(a) Simulate the actual service mounting (b) Cause no dydaEic coupling to the test item.
- 8. The in-s,itu application of vibratory devices to superimpose the >
seismic vfBrat'ory loadings on the complex active device for operability testing is acceptable when application is justifiable.
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- 9. . The test program may be based upon selectively testing a repre-sentative number of mechanical components according to type, load level, size, etc. on a prototype basis.
II. Seismic Design Adequacy of: Supporto
- 1. Analyses or tests shobid be performed for all supports of electrical and mechanical equipment and instrumentation to ensure their structural capability to withstand seismic excitation. j 4
- 2. The analytical results must _ include the following- l (a) The required input motions to the mounted equipment should be obtained and characterized in the manner as stated in Section I.2. l l
l (b) The combined stresses of the support structures should be l within the limits of ASME Section III, Subsection NF - )
" Component Support Structures" (draf t version) or other l comparable stress limits. l l
- 3. Supports should be tested with equipment installed. If the l equipment is inoperative during the support test, the response '
at the equipment mounting locations should be monitored and characterized in the manner as stated in Section I.2. In such a case, equipment should be tested separately and the actual input to the equipment should be more conservative in amplitude and ! frequency content than the monitored response.
- 4. The requirements of Sections I.2, I.4, I.5, I.6 and I.7 are ;
applicable when tests are conducted on the equipmen; supports. '
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4.0 Reactor 4.78 Discuss the empirical and/or analytical methods applied to fuel clad mechanical interaction problems discussed in Section 4.2.1 including cladding strain due to fuel swelling, fuel cracking or pellet edge effects, radial and/or axial differential thermal expansion of the fuel and cladding.
Your response should provide enough information to permit us to establish that the cladding stress or strain criteria are satisfied during steady state, transient, and accident conditions.
a 4.79 Provide the analytical, test, or empirical basis to demonstrate that the fuel channel box can maintain its dimensional integrity, strength and spatial position throughout its lifetime. The answer should consider steady state, transient and accident conditions.
4.80 Provide the seismic analysis used to evaluate the Fuel Assembly.
This should include:
a) mathematical model, b) numerical results, c) a discussion of the safety margins, d) justification for the assumption that the channel box can assume all the load, e) a justification to demonstrate that the assumption that a very simple mass spring system may be used to represent a whole fuel assembly in the core can provide realistic stress and deflection values, and f) a discussion of fluid sloshing effects en a fuel assembly during seismic response.
4 . 6 4.81 Present the blowdown force analysis for the fuel bundle and channel box for the loss of coolant and steam line break accidents. Discuss how the thermal shock during depressurization and quenching affects the fuel cladding and fuel assembly integrity. Discuss the analytical methods used to determine fuel bundle and fuel assembly response due to these pressure loadings. List stress levels calculated for the fuel assembly's-structural components. ,
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4.82 Provide a detailed description of the methods and procedures
used in the dynamic system enalysis which will be preformed to confirm the structural design adequacy of the reactor internals d (including fuel element assemblies, control rod assemblies and drives) to withstand dynamic effects under a simultaneous occur-
. rence of steam line break and SSE.
4.83 Provide a summary of the safety analysis of the fuel assembly performance in terms of a stress, report for each component and loading category.
An example is given in Attachment 1 to Section 4. For this phase of our review, a listing of the allowable values is accentable.
4.84 Provide a description of the models used in the three-dimensional boiling water reactor simulation code that is referred to in Section 4.3.3.1, and in Section II of GESSAR Appendix 4A.
- 4. P5 In Section 4.3.2.2, and on pp. 46-47 of GESSAR Appendix 4A, it is stated that one or more power distributions are selected (" design power distributions") which are more limiting than expected operating conditions, and that it is then verified that under these more stringent conditions, the design limits are met. The design power distributions i
utti fnem the basis for the Tech Specs. This should be accounted for in the plant design and in the following questions:
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- 4. 85.1 What design (or standard) power distributions are used in addition to the one shown in Figure'20a of GESSAR Appendix 4A?
- 4. 85.2 What are the procedures used'to demonstrate that the design power distributions represent the thermally most limiting conditions expected during the. cycle?
- 4. 85.3 What application is made of design power distributions in defining beginning condition:a for transient and accident analyses?
4, 86 Explain the footnote referring to the water rod in the proprietary [
l Table 4.3.6, provided in response to question 4.50. ;
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Attachment' I to Section 4 '
An Example of Stress Report Fuel Rod Water Rod Channel Others x 1 30,000 psi 30,000 40,000 w
.8 2 2 20,000 5,000- 10,000 etc.
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- x h 1 50,000 psi 50,000 60,000 5 w o C 28 2 40,000 20,000 55,000 etc.
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" " 3 1.35 2.5 1.00
$ c 1 x% x' %
0.' Y y% y' % etc. etc.
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u it. 1 60,000 psi 60,000
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1.33 1.20 M l3 -
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- t. t e o e E " " 0.2 0.01 0.6 EE u o Note: 1. allowable value
- 2. calculated value
- 3. safety margin
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i 5.0 Reactor Coolant System ar.d Connected Systems 5.25 Pages 5.5-12 and 5.5-13 state that based on the analysis of the main steam ~
line isolation valve failure in the Duane Arnold FSAR application a main steam line seal system is not required. It is our position that a sealing system will be required to supplement the isolation function of the main j steam line isolation valves. Therefore, provide sufficient information to demonstrate that the GESSAR MSLIVSS is in compliance with the'following:
- a. The main steam lint-isolation valve sealing system (MSLIVSS) and any necessary subsystems should be designed in accordance with seismic Category I requirements, including the source of sealing liquid if ;
I a a liquid seal is used.
- b. The MSLIVSS and any necessary subsyst' ems should be capable of with-standing effects associated with (a) internally-generated missiles, (b) the dynamic effects associated with pipe whip and jet forces, and (c) normal operating and accident-caused local environmental conditions consistent with the design basis for plant safety systems.
- c. The MSLIVSS should be capable of perforcing 1ts intended function following any single active failure, including failure of any one of the main steam line isolation valves to close.
- d. The MSLIVSS should be capable of performing its intended function following a loss of all off-site power coincident with postulated accidents. !
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- e. The MSLIVSS should be designed with sufficient capacity and capability to achieve a positive seal of the main steam lines and preserve containment integrity under the conditions associated with ,
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postulated accidents, 1
- f. The MSLIVSS may be manually initiated and controlled and should be )
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O designed.to permit actuation in a' time period of about 20 minutes following postulated accidents. This time period should be consistent with loading requirements on the emergency electrical buses'and with .
reasonable times for operator information, decision,'and action.
- g. Instrumentation and controls necessary for the functioning of the MSLIVSS should be designed in 'accordance with standards applicable
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to nuclear plant safety-related instrumentation and control systems,
- h. The MSLIVSS controls should be provided with interlocks actuated from the reactor protection or containment isoittion systems to prevent inadvertent MSLIVSS operation.
a 1. The MSLIVSS should be designed to permit testing of the operability
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of controls and actuating d'evices during power operation to the extent practical and testing of the complete functioning of the system during plant shutdowns.
- j. The MSLIVSS should be designed so that effects resulting from a sealing system failure will not affect the integrity of the main steam lines or isolation valves,
- k. The MSLIVSS should be designed so thoc any effects result 1ng from use of a liquid sealing medium, such as taermal stresses, pressures associated with flashing, and thermal deformations under the loading conditions associated with the activated system, will not affect the structural integrity of the main steam lines and isolation valves,
-and that any deformation of isolation valve internals will be limited-to values that will not induce excessive leakage of the isolation l
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- 1. The MSLIVSS should include provisions to prevent or control valve steam packing leaking or other direct leakage from main steam line isolation valves outside containment.
5.26 The response to-Question 3.73 is not complete. Provide the information for Class 1 field run piping systems in Section 5.2.1.19 that is provided in Section 3.9.2.7 for Class 2 and 3 field run piping systems.
5.27 Supplement..thagrggponses.to Questions 5.3 and 5.17, by adding the program PDA-Pipe Dynamic Analysis Program for Pipe Rupture Movement (NEDE-10813) l l
t'o the list of' programs. In addition, provide the information requested a
in the attachment to Question 5.3 entitled " Acceptability of Computer Programs Analysis of Mechanical Components and Equipment." I 5.28 Provide your program for the inservice testing of pumps and valves that are part of ASME' Boiler and Pressure Vessel Code,Section III, Class 1, 2, and 3 safety'related systems. Furnish sufficient information to indicate that the pump test program meets the provisions of ASME Section XI, Summer 1973 Addenda, Subsection IWP, and that the valve test program meets Subsection IWV.
We will require that the design criteria, at the preliminary design stage of review, of the affected systems should provide for access to and location of components in such a manner as to reflect the test program requirements.
5.29 Your position with respect to the use of Type 304 s ainless steel (Reg. Guide 1.44) in GE BWR's is acceptable provided that an inservice inspection program, to include PT and UT examination of the HAZ of welds, be made once each 10-year inspection interval on all removable reactor I
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f internals fabricated of Type 304 stainless steel. Propose a test that you believe is meaningful for determining the degree of sensitization of the heat af fected zones of qualification welds.
5.30 The requirements of 10 CFR 50 Appendix G must be met in Section 5.2.4.6.5.
5.31 It is our position that the predicted shift in RT presented in Section NDT 5.2.4.6.8 of GESSAR is acceptable only for those reactor vessels purchased to the restricted copper and phosphorus contents for beltline materials shown in Table 5.2.4.6.8-3).
5.32 Section 5.2.4.6.9 discusses how the surveillance capsules are positioned in the vessel. If the design of the capsule holders <
is the same as you have used on previous GE designed plants, provide attachment details of the capsule holder brackets-to-vessel welds. If you are planning a new design in this area for GESSAR, provide a schedule of when this information will be available. Our concern is that there have been reports of corrosion under cladding that has weld overlay.
5.33 It is our position that fabrications of welded austenitic stainless steel Classes 1, 2, 3, and CS components should comply with the requirements of Section III and Section IX of the ASME Boiler and Pressure Vessel Code supplemented by the following:
5.33.1 The veld filler materials shall meet the following additional requirements:
- a. Delta ferrite determinations should be performed on undiluted weld deposits f or each Lot and licat of A-8 weld netal (Para. QW 442 of Section IX), except that delta ferrite determinations wi.1 not be required for SFA-5.4 Type 16-8-2 wcld metal, nor for A-S weld Filler
metal to be used for veld metal cladding. Delta ferrite determinations for consumable inserts, electrodes, rod or wire filler metal used with the gas tungsten are process, and deposits made with the plasma arc welding process shall be predicted using an applicable constitution diagram,1 to demonstrate compliance with the Regulatory position for the amount of delta ferrite as shown in Position 1(b) of Regulatory Guide 1.31.
- b. For delta ferrite determinations made on undiluted weld deposits.
the weld pad employed shall be made and tested in conformance with the applicable sketch and methods described in the American Welding
'd Society Specification SFA-5.4. The undiluted weld deposits should-contain between 5.and 20 percent delta ferrite.
- c. Chemical analysis be performed on undiluted weld deposits.
1 5.33.2 The results of the destructive and n6 destructive tests required in 5.33.1 above should be included in a Certified Materials Test Report as required by ASME Code Section III, NB-2130 or NB-4130.
5.34 Acceptable alternatives to positions 3 through 7 of Regulatory Guide 1.31 are given below.
5.34.1 Production welds should be examined to verify that adequate delta ferrite levels are present, by magnetic measurement devices. Welds 1 inch or greater in th1ckness shall be examined on a 100% basis, whereas, a sampling plan may be used for examination of welds less ,
I than 1 inch in thickness The examination should show that each weld contains at least 3* delta ferrite based on the sverage of four test l l
readings taken on the face of the completed weld deposit at the Schaeffler, Modified Schaeffler, or DeLong Diagram, American Society for Metals Handbook, Vol. 6, pp. 246-247.
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l- centerline of the weld, and at 1/4 weld length intervals. Instrument l
readings should not be taken at " starts and stops" locations, or in l weld beads adjacent to the base materials. The average of four instrument readings should not include any reading below one percent delta ferrite. Weld locations that show 1% or less delta ferrite may i
be reexamined.
The magnet'ic instruments used for examination of the weld pads for delta f errite should have been calibrated using secondary standards traceable to National Bureau of Standards
- standards, and to a Magne-gage using the procedures shown in Welding Council document of July 1, a '
1972, " Calibration Procedures for Instruments to Measure the Delta Ferrite Content of Austenitic Stainless Steel Weld Metal," and as I supplemented by procedures shown in AWS specification A 4.2-74.
5.34.2 The upper limit on delta ferrite shown in 5.33.1(b) above need not be required for welds that do not receive heat treatment subsequent to welding.
5.35 In the event that 5.34.1 above is not met, the' respective production welds may be evaluated for acceptability using:
- a. An analysis of service requirements of the weldment, and comparison of these requirements with the ASME Code criteria for fatigue strength, but using conservative fatigue data that accounts for veld metal with micro-fissures, or
- b. An examination of the weld or welds to demonstrate the absence of unacceptable fissures or cracks. Where the production weld is below the minimum acceptable level of delta ferrite, a sample of the weld
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' ' ' . . 1 shall be removed, and a metallographic examination or a bend test shall be made on a transverse section to determine the presence i
or absence of excessive' fissures.. The' acceptance criteria is as
! .follows: Fissures 1/64" and less shall not be counted. The presence of a single tear or fissure larger than 1/16", or of a greater number f
.than 3 of size between 1/64" and 1/16" in any 0.2 square inch of weld- l i
- metal. .shall constitute failure of the test.
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- c. Welds found unacceptable shall be repaired .and reexamined for delta ferrite content in accordance with the procedure shown in 3(a) above.-
a An' example of a suitable examination plan is discussed in Attachment 1 to Section 5.
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Attschmint No.1 '
to Section 3 EXAMPLE OF A SUITABLE. STATISTICAL SAMPLING PLAN:
'A.. Examination (See Table I)
- 1.'; Production Welds Creater Than One (1) Inch in Thickness:
Delta ferrite determinations will be made on the completed surface
, of all such welds. When observed average delta ferrite levels of ,_
3% or more are indicated, all of the velds will be considered acceptable, and no further testing is needed.
'In the event that the delta ferrite level in some welds is lower than an average,of 3%, a metallographic examination or a macroscopic examination earformed on transverse' side-bend specimens to determine.
a the presence of microfissures will be made. The specimens for metallographic or macroscopic examination will be selected'from the welds' exhibiting delta ferrite levels lower than 3% average.
in accordance with Table 51.
- 2. Production Welds One Inch or Less in Thickness: (See Table I)
Delta ferrite determinations will be made on the completed surface of such welds, selected in accordance with' Table II. If observed delta ferrite levels of 3% or more are indicated for. the sample welds, (i.e., Column 2, Table II) the entire batch of welds (i.e., Column 1),
that the sample represents shall be considered acceptable. If the number of welds in the sample size having.less than an average of 3%
delta ferrite equals or exceeds the rejection level (Column 3), all the welds in the batch (Column 1) will have to be inspected.
For velds having an average delta ferrite level less than 3%, a macro-scopic examination will be performed on trahsverse side-bend specimens to determine the presence of microfissures, or metallographic examina-tion will be made on specimens cut from the velds. Specimens for such l
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. Attachmint No.1 examination.will be selected from welds exhibiting delta ferrite levels lower than an average 3% in accordance with Table II.
- 3. Sample. lots or batches for velds greater than one inch in thickness, and for welds one incb or less in thickness will not be grouped ,
l together, and will be macroscopically examined on a separate batch
- basis.
B. Acceptance and Rejection for Delta Ferrite Levels and Microfissures
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(All production welds)
- 1. All welds having an average delta ferrite level of 3% or more are acceptable.
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- 2. Welds having less than 3% average de'Ita ferrite shall have transverse.
side-bend tests taken and the welds shall be examined macroscopically for fissures, or the welds shall be examined metallographically, with sampling to be in accordance with Table II. Microfissuring1 detected during these inspections shall be evaluated by the following criteria: Fissures 1/64" and less shall not be counted. The presence of a single tear or fissure larger than 1/16", or of a greater number than 3 of a size between 1/64" and 1/16" in any 0.2 square inch of weld metal, shall constitute failure of the test.
- 3. Welds found unacceptable shall be repaired and reexamined by the above procedure.
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. . Attachment No. 1 TABLE I No. of Welds Welds having Welds having less.
Thickness of to be magnetic average 3% or more than average 3%
Weld Inspected delca ferrite delta ferrite one (1) inch all welds OK - no further Inspect for fissures or greater. examination per Table II necessary Less than one Inspect on a If all of sample Inspect for fissures
.(1) inch sampling basis batch are OK no 'per Table II.
per Table II ,further , exam.
required for entire group TABLE II Column 1 Column 2 Column 3 Total No. of . Sample Size - No. of Welds Welds to be examined Rejection Level
- 2-8 2 1 9-15 4 1 16-25 6 2 26-50 10 2 51-90 16 2 91-150 26 2 151-280 40 4 281-500 64 5 501-1200 100 7 1201-3200 160 9 3200-10,000 250 13
- If the welds examined and found unacceptable reach the figure shown in Column 3, the welds shown in the representative batch in Column 1 shall
. be rejected.
[6.0.. . Enginsarad Saf aty Fatturse 6.89 Question' 6.45' on engineered safety feature air filtration systems is j
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'not answered adequately. Each position in Regulatory Guide 1.52 should be addressed. A simple statement of compliance with the Guide'is insufficient. Therefore, analyze each engineered safety feature air filtration system (ECCS room, control room and SGTS)
I as to the positions in Regulatory Guide 1.52, " Design, Testing and Maintenance Criteria for Atmosphere Cleanup System Air Filtration and Adsorption Unitu of Light-Weter-Cooled Nuclear Power Plants."
Provide a tabular listing showing each engineering safety feature air filtration system and.its compliance or non-compliance with each position in Regulatory Guide 1.52. Each item of non-compliance should 4)e thoroughly explained.
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. . 8.0 Electric Power 8.12 In .regards to the diesel generators the following statements have been noted:
- a. Design of the overall onsite a-c power systems is the responsibility of the applicant. Development of that portion within the nuclear island is the responsibility of General Electric (page 8.3-1).
- b. From Figure 1.2-1 General Arrangement Site Plan, it would appear that the two buildings housing Division I, II and III diesel generators are within the nuclear island.
- c. Figure 1.2-3 as well as Table 9.2.2 appear to indicate that Division I and II loads will be powered by General Motors tandem engine driven generators.
- d. The footnote on Table 9.2.2 indicates that procurement of Division I and II diesel generators will be the applicant's responsibility.
8.12.1 From the above it is not clear if the diesel generators supplying power to Division I and II loads are within the scope of GESSAR. Clarify whether the diesel generators are within the GESSAR scope of supply. If they are provide further information concerning their design and the design of their auxiliaries such as the lube oil cooling system, etc.
8.13 The concerns expressed in Question 8.6 relating to the location of the diesel generator air intake and exhaust were made as a result of the l
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original Figures 1.2 -5 and 1.2-8. These figures have been revised and now omit the intake and' exhaust information. Provide figures showing the location of the diesel generator air intake' and exhaust and discuss the concerns raised previously in Question 8.6.
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[ 9.0 Auxiliary Systems l 9.39~ In reference to the fu&1 pool cooling system on page 9.1-14, the statement is made "The fuel pool cooling system shall be classified seismic Category I to ensure cooling of the spent fuel in case of i-
.a seismic event. Table 3.2.1 appears to indicate that the vessels containing the parallel filteY/demineralizers (which are shown in Figure 9.1-3b as being in series with the pumps and heat exchangers) do not meet seismic Category I requirements and.therefore, are in conflict with the statement on page 9.1-14. Clarify this inconsistency.
9.40 Section-9.1.3 states that th'e fuel pool cooling and cleaning. system a has been sized based on the following:
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-(a) The maximum normal heat load will consist of the decay heat from an average equilibrium fuel cycle batch of spent fuel just discharged from the core plus the decay heat from a batch that has decayed from the previous refueling, which has been in the pool for one year.
(b) The fuel pool water temperature will be maintained at or below 125'F when both A and B fuel' pool cooling loops are in operation.
The above heat load is not the actual maximum heat load that the pool may be required to handle at some point over the life of the facility. The maximum possible heat load is provided in Section 9.1.3.3. To accommodate these larger heat loads, the heat removal capability of the RHR systems will be required.
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In order to provide sufficient assurance that a situation is not.
possible where the RHR system is required for both fuel pool ,
cooling and ECCS, it is our position that the reactor will be shut-down should either the fuel pool water temperature exceed 150*F or I whenever portions of the RHR syste.m are required to assist in cooling the fuel pool water.
9.41 Indicate if the four components, listed in the table given in
! response to question 9.20.1, are within the scope of GESSAR. If 1
l' they are part of GESSAR, provide the design data for these Components.
- 9. 4"2 The response to question 9.20.3, indicates that the Fuel Pool Cooling and Cleanup System and the CCW system P&ID's will be
' modified . These revisions should provide sufficient information to give the reviewer assurance that any one f ailure in an essential or non-essential system can not interact in such a fashion as to degrade or disable an essential water system.
Indicate all interfaces that exist, between GESSAR and balance of plant, on all of these water systems.
9.43 The response to 9.21 discusses the demineralized makeup water and condensate storage systems and their intermittent flows to certain nuclear island systems, such as the RHR system flush, which are instrumental in attaining and maintaining a cold, safe shutdown.
Discuss the criteria followed by GESSAR or imposed on the balance of plant design to provide assurance that any single failure, valve malfunction, or improper valve position in these systems will not degrade or disable a water system essential to the facility and thereby prevent the attainment of a safe shutdown.
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. . 9.44 Provide a P'and ID of the instrument and service air system within the scope,of GESSAR. The diagram should show all major components, t
within the nuclear island such as compressors, receivers, pressure regulators, air treatment devices such as dryers and filters, valving, piping, interties between the two systems as well as all' safety related components and other significant components processing or utilizing compressed air. Relevant design data such as capaci-ties, design and operating pressures should be included on the diagram. ,
9.45 In. response 9.25 the following statement is made "No intertie of
, the instrument air kystem and the service air will exist within the Nuclear Island." Assuming the above statement is correct, explain how the safe design bases (1) "During accident conditions,.
service air use will be restricted in order to conserve the instrument air supply pressure" can be true.
9.46 It is noted that the Radwaste Building Ventilation System has
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been revised to eliminate som'e areas being under positive pressure relative to non-contaminated areas.
Revised drawing (Figure 9.4.3) indicates no path for air flow between non-contaminated areas and contaminated areas in the radwaste building.
9.46.1 What method is planned to replace the air that is exhausted from the contaminated area?
l l 9.46.2 The referenced figure indicates an unobstructed path for exhaust l
'$ 8-air. originating in the contaminated areas and ducted directly to the plant vent. What provisions have been made to cleanup the exhaust air?
l 9.47- Your response to question 9.38 is not acceptable. It is our position that continuous purging of the containment is unaccept-able and internal filters are required for containment atmosphere cleanup. Provide the safety design bases for this system'as well as the preliminary design, if available.
9 48 The control room once-through emergency charcoal filter discussed in Section 9.4.lis not suf ficient to meet Criterion 19. Previously submitted questions 9.37 and 15.37 through 15.37.4 should be reviewed and a filter size and a
system design selected based on the results. Based on current information, GESSAR does~not meet GDC 19.
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. 11.0 Radioactive Waste Management l
11.24 Your response to question 11.22 referenced an August 6,1973, submittal on topical report NED0-10751,. Even with this information your report is not complete since it does not contain data regarding the performance of large
} scale systems. You are currently conducting tests on a low temperature large scale charcoal delay system at the KRB reacotr, which we consider to be the type of data necessary to support your design values. Provide us with a timetable for transmitting the results of these tests to us.
- 11. 25 In response to question 11.14, you stated that you considered it impractical to verify the absence of free water in drummed solid waste. We consider this response unacceptable. It is our position that this verification should be made and consideration should be given to such methods as ultrasonic and electrical conductivity for making this determination.
11 26 The design of the liquid radwaste system described in Section 11.2 of GESSAR does not provide the capability to automatically terminate liquid radwaste discharges in the event of high radioactivity levels in the discharge line.
The system does not provide for controlled release of liquid effluents, as required by General Design Criterion 60 of Appendix A to 10 CFR Part to. It is our position that the capability to automatically terminate radioactive liquid releases should be included in the design.
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> e 12.0 Radiation Protection 12.59 Your description of the radiation ources given in Section 12.2.3 is incomplete in that it does not consider the consequences of relief valve actuation to the suppression pool, and the resulting radiation exposure to operating personnel who could be in the containment at that time. Provide the results of an analysis of the frequency and duration of relief valve openings and the estimated radiation' exposure to operating personnel. For an incident such as this, the appropriate source term should be based on an offgas' rate of 350,000 pC1/see at 30 minutes decay in addition to ' activation gases. State and justify all other parameters and assumptions used in your analysis.
12.60 It response to question 12.47, you state that the design criteria for the effluent monitoring system would be determined on an individual plant basis.
It is our position that, for plants employing standardized GESSAR radwaste management systems, the design criteria for monitoring 'of ef fluents are I
sufficiently similar that they should be included in the standard plant design.
Provide a table giving the locations of monitors, and their control functions necessary to meet the requirements of General Design Criteria 60 and 64.
Provide the radionuclides and expected concentrations during normal operation 1
including anticipated operational occurrences. l 4
l 12.61 Indicate the precautions taken to insure that electronic equilibrium was obtained j 1
in the PIC-6 ionization chambers used and referenced on page 12.1 -12.
12.62 Page 12.1 - 20 of GESSAR indicates that the average personnel exposure is )
estimated to be 240 man-rem per year. Provide the detailed basis for this 1
estimate since data we have received from operating reactors indicates exposure '
may be higher than ' estimated.
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12.63 The response to question 12.23 is unacceptable. Provide the maximum source 1
L strengths used in the layout and shielding design of pipe chases.
l- 12.64 . The response to question 12.38 is inadequate. " Field run piping" is piping that is layed out and physically p1'ced a in whatever locations are conveniently available by the on-site piping contractor. In order to meet the design guides I listed in Section 12.1.2.1, it is our position that " guidance" provided to construction engineering must include the specific constraints on the layout of field run piping carrying radioactive materials.
12.65 Specify now that information obtained from the AIF study of occupational exposure will be factored into the GESSAR design (as stated in the response to question 12.48.
12.66 Response'ipcomplete. Provide detailed information indicating how the area above the fuel pool will be exhausted such that operator exposure to air-borne radioactivity will be minimized.
12.67 Section 12.2.4 and the response do question 12.55 indicate that the air-borne radioactivity monitors in the exhaust system f rom the Fuel Building, Reactor Building, Radwaste Building and Auxiliary Building will ensure that personnel are not subjected to airborne radf oactive contamination in excess of those limits in 10 CFR 20, Appendix B Table I. The system ,
e described is primarily an ef fluent monitoring system and is not adequate for ;
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measuring airborne radioactivity to meet the requirements of 10 CFR 20.103 and 20.201. It is our position that sufficient fixed airborne radioactivity ]
1 monitoring be provided to assure the requirements of 10 CFR 20.103 and 20.201 are met. The use of portable equipment for airborne monitoring, although necessary, cannot be considered a substitute for continuous monitoring.
12.68 The response to question 18.24 indicates that improper design of the turbine i
building could affect the dose rates within the nuclear island. Indicate the j radiological interface constraints that need to be imposed on the BOP design
to assure that the doses to personnel are not increased as a result of-improper BOP design.
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15.0 ' Accident Analysis 15.42 The response to question 15.41.does not adequately answer our concerns.
All estimates of the radiation levels shown in Table 3.11-3, for which equipment must be designed as a result of design basis accident should include 50% of the halogens (iodines plus bromines), 100% of the noble gases, and 1% of the particulate. The doses should include an estimate of the beta ray doses.for all exposed organic materials, as well as the gamma ray and neutron exposures. D'iscuss the-bases for the calculation of the LOCA integrated doses in the drywell and for the standby gas treatment system. Provide the source' terms in each region as a function of time and energy, describe the methods used to compute the doses a (listing any computer codes used), and state all other assumptions used.
15.43 Provide the dose contribution following a loss-of-coolant accident due to leakage past the main steam line isolation valves prior to actuation of the r.equired main steam line sealing system assuming that all valves are leaking at the Technical Specification leak rate and the Containment Drywell a tmosphere begins to leak past the outboard valves at the time the accident occurs. State all assumptions.
15.44 Your. response to question 15.40.3 did not include the limiting value of
.X/Q acceptable for sites using the GESSAR plant. Provide this value.
Our preliminary calculations, based on plant design details you presented in GESSAR, indicate a 5 percentile meteorology for the 0-2 hour period of less than 8 x 10 -6 sec/m in order for a site to be acceptable for a GESSAR plant. That figure was derived on the assumption that, in the absence of an adequate response to Question No.15.38, no credit can be allowed for mixing or holdup in the secondary containment.
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L 15.45. -With. regard to your response to question 85.35, we question the applicability-of steady state halogen carryover data in determining the percent carryover of halogens released from the fuel elements as a result of a rod drop accident. Entrainment of halogens in non-condens ble clad gap gases or in steam voids or bubbles that are carried up to and through the water surface is possible. This could result in higher carryover than predicted from steady state operation where steam and non-codengible voiding are not significant carryover mechanisms.
l It is our position that 10% carryover provides a more ' reasonable margin i to account for these entrainment effects. In this regard provide dose computations using the 10% carryover assumption.
15.46 For the design basis loss-of-coolant accident description in Subsection
- 15. 1. 39 .3, include a chronology of events, such as, when containment purge lines are isolated, when the SGTS starts (relative to when- the LOCA occurred) and at what flow rate, and when the recirculation dampers I close in the shield building exhaust and recirculation system. The data in Tables 6.2.4 and 6.3.1 apply to the containment response and ECCS operation only and do not include operation of fission product control systems.
15.47 The response to question 15.38 is inadequate. The description and drawings of the Shield Building Annulus Recirculation and EWaaust System provided in Section 9.4.5.3 indicate that following a loss of coolant accident this system will be used for exhaust only. Therefore, no mixing bredit !
will be allowed in the staff's model for the LOCA dose calculation. ?
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leakage which goes to the filtered areas maintained at a minimum negative pressure differential of -0.25" water gauge and '(2) no mixing in the annulus.
15.48 In the response to question 15.38 you also state that leakage bypassing the shield building annulus will go into areas which are maintained at a -
negative pressure and vented to the SGTS. On Table 6.2.9 identify in which of these vented areas each penetration terminates. Include a table listing personnel air locks, equipment hatches, and all process lines which have guard pipes open to containment or drywell atmosphere <
"following a LOCA. Identify the vented area each guard pipe terminates k outside containment. Leakage from any of these paths which could be f rom containment to non-treated volumes is to be treated as bypass leakage in the LOCA dose analysis. For penetrations which are tested to zero ,
leakage, the test sensitivity should be included' as bypass leakage.
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