ML19261B162
| ML19261B162 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 01/29/1979 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Jens W DETROIT EDISON CO. |
| References | |
| NUDOCS 7902140104 | |
| Download: ML19261B162 (29) | |
Text
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UNITED STATES y p, e f j> j NUCLEAR REGULATORY COMMIhbbOh ~ ~ # - - - -..: 4
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W ASHINGTON, D. C. 20555
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J/JJ 2 3 G73 Docket No. 50-341 Dr. '4ayne H. Jens Assistant Vice President Engineering & Construction The Detroit Edison Company 2000 Second Avenue Detroit, Michigan 48226
Dear Mr. Jens:
SUBJECT:
REQUESTS FOR ADDITIONAL INFORMATION IN FERMI 2 FSAR As a result of our continuing review of the Final Safety Analysis Report (FSAR) for the Enrico Fermi Atomic Power Plant Unit 2, we have developed the enclosed requests for additional information.
Appendix 10 A to the FSAR, " Steam Bypass System" filed in Amendment 19 on January 16, 1979 does not include the method of analysis for the turbine trip transient. The information needed for our review of this transient is contained in the NRC memorandum dated December 27, 1978,
" Summary of November 28, 1978 Meeting Regarding Turbine Trip Transient Analysis" and was discussed by telephone with Mr. L. Schuerman on January 19, 1979.
Please amend your FSAR to revise Appendix 10 A and to comply with the requirements listed in the enclosure. Our review schedule is based on the assumption that the additional information will be available for our review by February 26, 1979.
If you cannot meet this date, please inferm us within 7 days after receipt of this letter so that we may revise our scheduling.
Sincerely,
/
(JohnF.Stolz,Chie tight !4ater Reactors Branch No.1 Division of Project Management
Enclosure:
Requests for Additional Information cc w/ enclosure:
See page 2 s
790214O WS
Dr. Wayne H. Jens Jr.N 2 0 1073 cc: Eugune B. Thomas, Jr., Esq.
Mrs. Martha Drake LeBoeuf, Lamb, Leiby & MacRae 230 Fairview 1757 N. Street, N.W.
Petoskey, Michigan 49770 Washington, D. C.
20036 Peter A. Marquardt, Esq.
Co-Counsel The Detroit Edison Company 2000 Second Avenue Detroit, Michigan 48226 Mr. William J. Fahrner Project Manager - Fermi 2 The Detroit Edison Company 2000 Second. Avenue,48226 Detroit, Michigan Larry E. Schuerman Licensing Engineer - Fermi 2 Detroit Edison Company 2000 Second Avenue Detroit Michigan 48226 Charles Bechhoefer, Esq., Chairman Atomic Safety & Licensing Board Panel U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Dr. David R. Schink Department of Oceanography Texas A & M University College Station, Texas 77840 Mr. Frederick J. Shon Atomic Safety & Licensing Board Panel V. S. Nuclear Regulatory Commission Washington, D. C.
20555 Mr. David Hiller University of Michigan Law School Hutchins Hall Ann Arbor, Michigan 48109
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Mr Jeffrey A. Alson 772 Green Street, Building 4 Ypsilanti, Michigan 48197
ENCLOSURE REQUESTS FOR ADDITIONAL INFORMATION ENRICO FERMI ATOMIC POWER PLANT UNIT 2 DOCKET NO. 50-341 Requests by the following branches in NRC are included in this enclosure.
Requests and pages are numbered sequentially with respect to previously transmitted requests.
B ra nc_h_
Page No.
Materials Engineering Branch - Materials 121-4 Integrity Section thru 121-8 Auxiliary Systems Branch - Fire Protection 021-1 Review thru 021-20 Reactor Systems Branch 212-21 (This oage supetsedes Page 212-21 transmitted January 3, 1979)
121-4 1 21.0 MATERIALS ENGINEERING BRANCH - MATERIALS INTEGRITY SECTION 121.10 The. 3AR states that compliance with Appendix G,10 CFR Part 50,'
and Appendix G of the ASME Code was not possible for components which were purchased prior to the issuance of the Summer 1972 Addenda. The applicant requested that the NRC staff evaluate the special methods of compliance on the basis of the last paragraph on page 19013 of the FEDERAL REGISTER, July 17, 1973.
We will require that the applicant explicitly identify the areas in which compliance to Appendix G of 10 CFR Part 50 is not possible. The applicant should provide the technical basis to demonstrate that an acceptable level of component integrity will be provided for those areas of non-compliance. The staff will evaluate the technical information in the responses and determine whether exemptions from the specific requirement of Appendix G are necessary and justified.
121.11 In relation to the fracture toughness requirements of Appendix G of 10 CFR Part 50 and Appendix G of the ASME Code,1971 Edition, including Summer 1972 Addenda, the chemical composition of the base plate and weld metals should be provided, and the governing material identified for the beltline region of the reactor vessel. We will require conformance to the recommendations of Regulatory Guide 1.99, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," in estimating the shift in the referenced nil-ductility temperature as a function of fluence.
121.12 We will require that the applicant explicitly identify the areas of non-compliance in the materials surveillance program to the requirements of Appendix H of 10 CFR Part 50. The staff will evaluate the program and determine whether exemptions from the requirements of Appendix H are necessary and justified.
In addition, the applicant should indicate areas of non-compliance in the materials surveillance program with the requirements of ASTM E 185-73, " Surveillance Tests for Nuclear Reactor Vessels."
121.13 We will require that your inspection program for Class 1, 2 and 3 components be in accordance with the revised rules in 10 CFR Part 50, Section 50.55a, paragraph (g) published in the Februa ry 12, 1976 issue of the FEDERAL REGISTER.
To evaluate your inspection program, the following minimum information is necessary for our review:
(1) A preservice inspection plan to consist of the-applicable ASME Code Edition and the exceptions to the code requirements.
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121-5 (2) An inservice inspection plan submitted within six months of anticipated commercial operation.
The preservice inspection plan will be reviewed to support the safety evaluation report finding on compliance with preservice and inservice inspection requirements. The basis for the determination will be compliance with:
(1) The Edition of Section XI of the ASME Code stated in your FSAR nr later Editions of Section XI referenced in the FEDERAL REGISTER that you may elect to apply.
(2) All augmented examinations established by the Commission when added assurance of structural reliability was deemed necessary.
Examples of augmented examination requirements can be found in NRC positions on (a) high energy fluid systems in SRP Section 3.2, (b) turbine disk integrity in SRP Section 10.2.3, and (c) feedwater inlet nozzle inner radii.
Your response should define the applicable Section XI Edition (s) and subsections.
If any examination requirements of the Edition of Section XI in your FSAR can not be met, a relief request including complete technical justification to support your conclusion must be provided.
The inservice inspection plan should be submitted for review within six months of anticipated commercial operation to demonstrate compliance with 10 CFR Part 50, Section 50.55a, paragraph (g). This plan will be evaluated in a safety evaluation report supplement. The objective is to incorporate into the inservice inspection program Section XI requirements in effect six months prior to commercial operation and any augmented examination requirements established by the Commission.
Your response should define all examination requirements that you determine are not practical within the limitations of design, geometry, and materials of construction of the components.
Attached are detailed guidelines for the preparation and content of the inspection programs and relief requests to be submitted for staff review.
e
121-6 GUIDANCE FOR PREPARING PRESERVICE AND INSERVICE INSPECTION PROGRAMS AND RELIEF REQUEST PURSUANT T0 10 CFR 50.55a(g)
A.
Preservice/ Inservice Inspection Program Description This program covers the requirements set forth in 10 CFR 50.55a(g) and the ASME Boiler and Pressure Vessel Code Section XI, Sub-sections IWA, IWB, IWC and IWD.
The guidance provided in this enclosure is intended to illustrate the type and extent of information that should be provided for NRC review.
It also describes the information necessary for " request for relief" of items that cannot be fully inspected to the requirements of ASME Section XI. By utilizing these guidelines, licensees can signifi-cantly reduce the need ~for having to respond to additional information requests from the NRC staff.
B.
Contents of the Submittal Tne information listed below should be included in the submittal:
1.
For each facility, include the applicable ASME B & P V Code date and appropriate addendum date.
2.
The period and interval for which this program is applicable.
3.
Include the proposed codes and addenda to be used for repairs, modifications, additions or alternations to the facility that might occur during this inspection period.
4.
Identify the examinations that you have exempted under the rules of ASME Section XI. A reference to the applicable paragraph of the code that grants the exemption is satisfactory. The inspect-ion requirements for exempt components should be shown; i.e.,
visual inspection during a pressure test.
5.
Identify the inspection and pressure testing requirem 7ts of the applicable Section XI requirements that are deemed impractical because of the limitations of design, geometry and material of construction of the components. Provide the information requested in paragraph C for the inspections and pressure tests identified.
C.
Reauest for Relief from Certain Inspection and Testing Recuirements It has been the staff's experience that many requests for r'elief from testing requirements subnitted by licensees have not been supported by adequate descriptive and detailed technical information. This detailed
121-7 information is necessary to document the impracticality of the ASME Code r quirements within the limitations of design, geometry and mater.als of construction of components and to determine whether the use of alternatives will provide an acceptable level of quality and sc'ety.
Relief requests submitted with a justification such as " impractical,"
" inaccessible," or any other categorical basis, require additional information to permit the staff to make an evaluation of that relief request. The objective of the guidance set forth below is to illustrate the extent of the information that is requitad by the NRC staff to make a proper evaluation and to adequately document the basis for granting the relief in the safety evaluation report. The NRC staff believes subsequent requests for additional information and delays in completing the review can be considerably reduced, if this information is provided initially in the licensee's submittal.
For each relief request submitted, the following information should be included:
1.
Identification the component (s) and/or the examination require-ment for which relief is requested.
2.
Number of items associated with the requested relief.
3.
ASME Code class.
4.
Identification of the specific ASME Code requirement that has been determined to be impractical.
5.
Information to support the determination that the requirement is impractical; i.e., state and explain the basis for requesting relief.
6.
Identification of the alternative examinations that are proposed in lieu of Section XI requirements or to supplclent partially performed Section XI examinations.
7.
Description and justification of any changes expected in the overall level of plant safety by performing the proposed alternative examinations in lieu of the ASME Section XI examination.
If it is not possible to perform alternate examinations, discuss the impact on the overall level of plant quality and safety.
For inservice inspection provide the following additional information
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regarding the inspection frequency:
121-8 8.
State when the relief request would apply during the inspection period or interval i.e., is the request to defer an examination.
9.
State when the proposed alternative examinations will be imple-mented and performed.
10.
State the time period for which the requested relief is needed.
Technical justification or data must be submitted to support the relief request. Opinions without substantiation that a change will not affect the quality level are unsatisfactory.
If the relief is requested for inaccessibility, a detailed description or drawing which depicts the inaccessibility must accompany the request. A relief request is not required for tests prescribed in Section XI that do not apply to your facility. A statement of N/A (not applic-able) or none will suffice.
D.
Recuest for Relief for Radiation Considerations Exposures of test personnel to radiation to acco1piish the examinations prescribed in ASME Section XI can be an importarit factor in determining whether or under what condition an examination riust be performed. A request for relief must be 335mitted and approved similar to that required for inaccessibility, We recognize that some of the radiation considerations will only be known at the time of th test. However, the licensee generally is aware, from experience at operating facilities, of those areas where relief is necessary and should submit as a min
- mum the following information with the request for relief:
1.
Total estimated man-rem exposure involved in the examination.
2.
Radiation levels at the test area.
3.
Flushing or shielding capabilities which t.ight reduce radiation levels.
4 Alternate inspection techniques proposed.
5.
Remote inspections considerations.
6.
Redundant systems or similar welds which can be inspected.
7.
Preservice and any inservice results of welds involved.
8.
Consequences if the' weld failed.
021-1 021.0 Auxiliary Systems Branch Fire Protection Review 021.1 Pace 98.3-5, Section 98.3.3, Electrical Reouirements_
You state in this section how various safety related cable trays and conduit are separated by distance from their redundant countercart and the criteria that was used to establish barriers between these redundant trains. A nore detailed fire hazards analysis should be conducted for each plant fire area, and the effects of a postulated fire involving permanent and/or transient combustibles (exposure fire) on systems, circuit cable trays, or equipment reouf red for safe plant cold shutdown should be evaluated.
In the fire hazards analysis you should identify all the redundant mechanical and electrical systems necessary for safe cold shutdown which are separated only by distance (no fire barriers). The fire hazards analysis should demonstrate that, assuming failure of the primary suporession system, a fire in installed and/or transient ccmbustibles will not damage redundant trains or divisions of systems reouired for safe olant cold shutdown. '4here this cannot be demonstrated, an alternate means of assuring safe plant shutdown (cold shutdown) should be provided.
021.2 cace 98.2-2, Section 98.2.2.1 and Pace 98.3-1, Section 98.3.1, Saf e Snutdown Cacabili ty You indicate that you are providing a remote shutdown canel for use in acneiving plant shutdcwn in the event of loss of the normal
021-2 shutdown cacabili ty.
In the event that both redundant trains of conduit-cable trays are lost due to a fire in the control room, or cable screading room, or remote shutdown panel room, it is our position that this control panel and its associated cabling b7 ccmoletely independent and separate of the normal shutdown systems. Describe how the alternate shutdown method will be carried out, state system modifications associated with the shut-down method, and confirm that written procedures have been established 'for immediate use by responsible individuals. The staff oosition with respect to providing alternate or dedicated shutdown methods for fire protection is stated in Enclosure 1.
Provide information that demonstrates that the recuirements of will be satisfied.
021.3 pane 98.a-1, Section 98.4, Hazards Analysis
'de notice that interactions between redundant trains of safety related eouioment and cable trays occur throughout the olant.
'dhere divisional interaction occurs, it is our position that:
A.
All interactions in the areas where the redundant safety related trains are within 20 ft. of each other should be identified.
Also, the consequence of electrically initiated or exposure fires should be evaluated w'th regard to plant shutdown capability (see Item 1).
S.
For those areas indicated in Ttem (A) above, where a fire can affect the plant shutdcwn cloability, an area automatic suo-pression system should be provided to afford protection against
021-3 exposure fires. Also, a 1/2-hr. fire rated barrier should be provided to separate one safety related train from the other or from a common exoosure fire source. Describe the type of fire retardant coating you will use to orotect cables at divisional cable tray interactions and at 0;her areas where caoles will be used.
C.
The suppression system should alarm and annunciate in the control room.
Section 98.4.1 Reactor Building 021.4 Pace 98.4-5. Section 98.4.1.3 Basement Corner Rooms, Zone 2. Ele-va tions $40'-0" and 562'-0" The basement corner rooms consist of four triangular shaped rooms, one of which is located in each corner of the reactor building.
Each room is comorised of two floors, or,a at elevation 540'and the other at elevation 562'.
An open stairwell in each room connects each flocr.
It is our position that the following be provided for each room on each elevation.
A.
A 3 hr. fire rated barrier be provided to separate each corner room from the torus room on each elevation where there are safety related equipment and cables.
3.
Due to limited access to each room, especially at the lowest elevation, automatic sprinklers should be provided on each elevation for each room, as the open stairvells of each room would make manual firefighting very difficult.
021-4 021.5 Pace 98.4-11, Section 98.4.1.6 First Floor, Zone 5, Elevation 583'-6" You are croposing to install a wet automatic sprinkler system over the railroad bay.
This area is open and adjacent to the floor area surrounding the drywell which contains safety related eovipment and cable from both divisions. The railroad bay also communicates through open stairways to safety related equipment and conduit / cable on the two elevations below, terminating on the 540' elevation of the reactor building. Considering a postuated fire with failure of the primary suppression system, it is our position that the folicwing be incorporated in the design of the plant:
A.
Stairways at coordinates B-17 and 3-10 should be enclosed in 2 hr. fire rated construction on the 583' elevation.
B.
A 3 hr. fire wall should be established along column line B from coordinates 9 to 17. All openings and penetrations should be properly protected to achieve the same fire rating as the wall.
021.6 Pace 98.4-15. Section 98.4.1.7 Second Floor, Zone 6, Elevation 613'-6" This zone encloses the entire floor area outside of the drywell with the exception of the RHR heat exchanger rooms. Since this floor communicates with the floor below and above, and because 5th redundant trains of conduit / cable are located in this area, it is our position that a wet pipe automatic sorinkler system be
021-5 provided to cover the entire floor area for protection against a possible exoosure fire of transient material.
021.7 Pace 98.4-20, Section 93.4.1.9, Fourth Floor, Zone 8, Elevation 659'-6" Describe hcw the drainage system is arranged around the lubricating oil in the couplings and cooling units of the two motor generator sets such that a fire in spilled oil in these areas will not spread through the drainage system to safety related areas of the plant. Verify that oil spills will not overflow the curbing provided for each hazard.
021.8 pace 98.4-24, Section 98.4.1.11 Drywell, Zone 10, Elevation 567'-0" to 684'-6" A.
Your fire hazards analysis for this area indicates that the drywell atmosphere will be inerted with a 97 percent concentra-tion of nitrogen. At a recent meeting you informed us, however, that you will no longer maintain this inerting environment in the drywell. Provide the folicwing additional information for this zone:
1.
Provide marked scaled drawings indicating the locating of the reactor recirculation pumos.
2.
We have been informed that you no longer intend to use Permali as a shielding material in this :ene. Verify that any shielding material used is of noncombustible construc-tion and revise the fire hazards analysis accordingly.
o e
021-6 B.
It is our position that the following be incorporated into the design of the fire protection system in order to ! evide a defense in-depth approach.
1.
Reactor Recirculation Pumps a.
Provide an automatic sprinkler or deluge system for each pump with alarm and annunciation in the control room. Adequate curbs and drains should be provided and discharged to a safe location to orevent such a fire
'from spreading to other areas, or b.
Provide an engineered completely self-contained oil collection system around all pressurized oil piping with proper drainage (including capacity) to a safe location.
2.
Install a permanent standpipe system inside containment with hose stations on each elevation.
3.
Smoke detectors (photoelectric) should be provided through-out the drywell area with alarm and annunciation in the control room.
Section 98.J.2 Auxiliary Buildino 021.9 Pace 98.4-37, Section 98.4.2.7 Second Floor, ? lscellaneous Rooms, Zone 6, elevation 613'-6" This zone consists of a welding eouioment area and personnel change roon. Since both safety related division I and II cable trays are not separated from these areas by 3 hr. fire rated construction:
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021-7 A.
Verify that the acetylene-oxygen fuel gas welding shop will not be located in this area. The welding shco should be located where it will not expose safety related equipment and/or conduit / cable to a ;otential fire.
B.
Install an autcmatic sprinkler system throughout the personnel change room, which has a high fuel loading, to protect the surrounding safety related ares from an exoosure fire in the personnel change room.
021.10 Dace 98.4-41, Section 98.4.2.10, Control Room, Zone 9, Elevation 643'-6" & 655'-6" It is our position that the peripheral rooms within the control room complex be separated from the control room by noncombustible construction having a fire resistance rating of at least I hr.
Each room should be provided with an automatic smoke detector which alarms and annunciates in the control room.
It is also our cosition that a cressure water type 2-1/2 gal. fire extinguisher be orovided for the control room complex for orotection against a deep-seated Class A fire.
021.11 Pace 98.4-43, Section 93.4.2.11, Division I and II Batterv Rooms, Zone 10, Elevation 643'-6" The battery room containing the division I batter'es is presently separated from its redundant battery room and surrounding areas by fire barriers of 1-1/2 hr. fire rating.
It is our position that the battery rooms be separated from each other and other are3s of the clant by fire resis_ tant construction of at least 3 hrs. Revise your fire hazards ana'fsis accordingly. Verify that loss of the
021-8 ventilation system for each battery room will alarm and annunciate in the control room.
021.12 Pace 95.4-52. Section 93.a.3, Residual Feat Removal Conclex The residual heat removal comolex contains the emergency diesel generator, diesel oil s*.orage tanks, RHR service water pumps, and various other safety related equipment. Train I is separated from train II by a 3 hr. fire barrier.
It is our position that you provide the folicwing:
A.
Due to the large volume of diesel fuel oil in the rooms adjacent to the emergency diesels, eso 3 hr. fire rated doors should be provided to protect all coenings in the fire rated wall including ventilation ocenings.
B.
Desc'.ce in detail the emergency drainage systems provided
'ar the diesel fuel oil as well as any firefighting water that may be used. Verify that there is no interconnection between division I and II via the drain system so that a fire will not spread from one division to the other through the drain system.
C.
Provide automatic smoke detection that alarms and annunciates in the control room in all areas of the Residual Heat Removal Complex.
D.
Provide an indeoendent alternate means of fire sucoression for each diesel fuel tank storage room. Loss of the primary suo-pression system. (automatic sprinklers) should not affect the functional cacability of the secondary fire suooression system.
w
021-9 E.
Verify that the volume of ^_.
individual fuel oil storage tank room is of sufficient capac1ty to hold 110% of the contents of its associated diesel fuel oil storage tank should a leak develoo.
F.
Describe how each tank is vented to the outside acd also what means are provided to prevent overflow into the building when the tank is being filled.
G.
Describe the type and fire resistance of the diesel fuel oil tank supports.
021.13 Pace 98.4-57, Section 98.4.5, Turbine Buildina The turbine building contains the RCIC pump discharge isolation valve E51F013 and HPCl pump discharge valve E41F006 which are located in the stear tunnel. Provide drawings of this safety related equipment indicating its relation to the area in which it is located showing all hazards and any fire protection provided. Also, provide structural drawings of the wall at each elevation of the turbine building where it abuts the auxiliary building and reactor building.
021.14 Pace 98.4-66, Section 98.4.8. General Service Water Pumchause Provide drawings, including elevations and sections of the general service water pumphouse showing the arrangement of all fire pumos, cioing and fire protection orovided.
021-10 Point-bv-Point Comoarison with Accendix A to NRC BTP APCSB 9.5-1 catea Auaust 23, 1976 021.15 Pace 93.5-6, Item 4, Sincie Failure Criterion Describe the provisions designed to prevent ligntning from initiating fires which could damage safety related equipment. Also describe what provisions have been made to prevent lightning from disabling the fire protection system.
021.16 Pace 98.5-7, Iten B, Administrative Procedures, Controls and Fire Bricade You have presented a_ descriotion of the administrative procedures you plan to utilize for fire prevention. We request that you re,1 w these procedures against the staff supplemental guidance contaiced in " Nuclear Plant Fire Protection Function Responsibilities, Admini-strative Controls, and Quality Assurance," dated June 14, 1977.
It is our position that you either:
(1) confirm that your existing administrative procedures and fire brigade and testing program meet the staff supplement guidelines, or (2) provide a comnitment that they will be revised accordingly.
021.17 Page 98.5-24, Item D.1, Buildina Desian A.
Verify that interior wa'l comoonents, thermal insulation materials, radiation shield.ag materials and soundproofing are noncombustible, and that int'rio ~ finishes are listed by a nationally recognized testing laboratory for flamespread, smoke e
0?l-11 and fuel contribution of 25 or less in their use configuration (ASTM E-84 Test, " Surface Burning Characteristics of Building Materials").
B.
Identify all areas where susoended ceilings are used and the criteria used for automatic detection and suopression systems where it was not practical to have these spaces devoid of combustibles.
C.
Substant. ate the fire resistance capabilities for the folicwing listed items as they pertain to safety related areas, areas exoosing safety related areas, or high hazard areas by veri-fying that their construction is in accordance with a particular fire tested design, and identify the design, test method, and acceptance criteria:
1.
Rated fire barriers including floor, ceiling and wall systems, structural members, and dv >rs.
Indicate the type of protective material used and the design number in reference to ASTM E-119.
2.
Fire dampers and fire doors including their method of installation in ventilating ducts penetrating fire barriers of safety related areas (fire door dampers are required in 3 hr. rated fire barrier penetrations).
For all instances where credit is taken for eauivalent rated docrs, it is our position that you provide a manufacturers credification for these doors.
0?l-12 3.
Fire barrier penetration seals around ducts, pipes, cables, cable trays, conduit, or any other openings. Verify that the seals are of the thickness specified in the tests and will meet the 3 hr. recuirements for ASTM E-119. Verify that the inplant cable tray supports are similar to the succorting arrangements used in the fire tests and that, in case of collapse of the trays, the resultant unsupported load and torque on the penetration seal will not affect the integrity of the seal. Those cenetrations in rated fire barriers (walls, floors, or ceilings) which are not sealed should be sealed in a manner providing a fire resistance rating equivalent to that of the barrier.
021.18 Pace 98.5-27, Item 0.2, Control of Combustibles Verify that piping systems containing flammable compressed gases or flammable licuids are not routed through or expose safety related equipment and/or conduit / cable.
021.19 Pace 98.5-30, Item 0.(f), Electrical Cable Construction Verify that electric cable constructions pass the current IEEE No.
383 flame test.
If the cables do not meet IEEE-383, provide test data, criteria, etc. of your Detroit Edison Soecification 3071-80 flame test and justify any deviation from IEEE-383.
e
021-13 021.20 Piae 98.5-31, Item 0.4, Ventilation 5
Cescribe the procedure employed for heat and smoke removal using fixed or portable equipment in areas that house safety-related systems or components. Describe how these areas can be ventilated for manual firefighting purposes. Consider that control or power cabling for normal ventilation may ret be functional in these areas.
Include a discussion regro e
control access to the equipment (including fire dampers,..
well as the ability to handle high temperature gases and particulates.
B.
Describe where the interlocks or local manual controls for resetting ventilation systems in areas protected by CO2 systems are located. Consider the effects of closed damoers ir hese areas, and. indicate how these dampers would be reset for venting the area for manual firefighting.
C.
In each area where safety-related systems or components are located, verify that products of combustion exhausted from the area will not be exhausted to other safety-related areas of the plant.
021.21 Pace 98.5-34, Item 0.5, Lichtina and Communication Verify that portable radio communication cnits will be provided at the site incorporating fixed repeaters as to acheive a safe shut-down in the event of a f'ce.
e
021-14 021 22 Pace 98.5-35, Item E.1. Fire Detection A.
Indicate whether tne fire protection detection system provided is a Class A or Class B system as defined in NFPA 720.
B.
State if the fire detection system circuitry is Class 1, per NFPA 70 or not.
If the fire detection circuitry is not Class 1, per NFPA 70, describe the physical separation criteria which assures that faults in the detection circuitry will not cause simultaneous fires in redundant safety division areas.
C.
Provide a detailed descriotion of the fire detection system, succorted where necessary, by diagrams or appropriate prints (include a single line drawing from the detection circuits, water-flow alarms, through the subpanels and into the control com).
Include an analysis, supported where necessary by test data, which substantiates that the sensitivity of fire detection devices and the number and placement of detectors are sufficient to provide detector response in time to prevent loss of safety-related systems or components. The analysis should include both fire detection devices used to notify personnel aM those used to activate fire protection systems.
D.
It is our positNn that crimary and secondary power be sucolied as follows:
1 Nor"al off site pcwer as the primary supply with a a hr.
battery suoply as the secondary supply; and 2.
Having capability for manual connection to an onsite source such as the_ Class IE emergency cower, or auto:ratically to a non Class IE source, which satisfies the requirements of NFDA 72D Section 2260, wT:hin a hrs. of loss of offsite oowe.
021-15 Such connection should follow the apolicable guidelines in Regulatory Guides 1.6,1.32, and 1.75.
Confirm that your design will meet this position.
021.23 Page 98.5-36, Item E.2. Fire Protection Water Sucoly A.
At present, thE prim &ry source of watr.r supply to the fire main system piping is provided by the p neral service water system pumps with the two fire pumps available as backup to maintain fire main system pressure. The connection to the fire main loop from the general service water system is isolable. However, it is our position that the fire main system be completely separate from service or sanitary water system piping.
B.
Verify that alams indicating pump running, driver availability and failure to start are provided for each pumo with alam and annunciation in the control room. Also, verify that the fire pump ar:d controllers are UL and/or FM approved.
021. 26 Page 98.5-39, Item E.3, Water Sprinklers and Hose Standpice Systems A.
At present, fire wrcer connect, ions are provided to the various but idings containing safety related eouipment and conduit / cable, however, both standpipes and automatic sprinklers are fed from the same connection for each buildinc. Headers fed from both ends are not provided inside these buildings. A single failure such as inadvertant valve closure can impair both the primary ar.d backup fire protection systems for these buildings. This arrar.gement is unacceptable.
It is our position that' modifica-tion be made to-the fire water system such that both primary and secondary water supply is not lost cue to single failure.
021-16 B.
It is our position that all sectional and divisional valves of the underground fire main be either locked open with administrative cortrols or be electricelly supervised with alarm and annunciation in the control room. We will also recuire all major fire protec-tion valves inside the building be eithEr locked open or electrically supervised.
C.
Verify that the 1 1/2 inch fire hoses are not more than 100 f t.
in length and that the interior manual hose stations will be able to reach any location with at least one effective hose stream.
D.
Provide scaled drawings showing the ccmplete fire protection under-ground fire main including all divisional, section and control valves as well as the placement of all hydrants, curb valves and hcse houses. Also provide an elevation drawing of the entire standpipe system including all valves and connections to the under-ground fire main.
021.25 Pace 93.4-43, Item E.6, Portable Extinauishers Indicate the type of portable fire extinguisher to be provided through-out the plant fore protection against a Class A deep-seated fire.
021.25 Pace 98.5 J7, Item F.2, Control Room You indicate here that the concealed space beneath the computer room floor will be provided with a total flooding Halon system. However, e
021-17 in the Fire Protection Analysis, Page 98.a-42, Section 98.a.2.10 ar.d on Drawing 6A721-2408 you indicate that this under floor space will be equioped with an automatic carbon dioxide suppression system.
Clarify this apparent discrepancy.
General 021.27 It is our position tha t fire stops be installed every 20 ft. along hori-zontal uncoated cable routings in areas not protected by automatic water systems. Between levels, or in vertical uncoated cable runs, fire stops should be installed at the midheight of the vertical run when its length is between 20 ft. and a0 f t.3 and at 15 ft. intervals in vertical runs of 30 f t. or more unless such vertical cable routings are protected by automatic water systems directed on the cable trays.
Individual fire stop designs should prevent the propagation of a fire for a minimum period of 30 min, when tested for the largest number of cable routings and mr.ximum cable der.sity. Confirm that your design will meet this position.
021.28 You have indicated that the ceiling in various areas of the plant, such as tne torus rocm is constructed of reinforced concrete over steel beams.
It is not clear from your description whether this steel is exposed.
If trese steel beams are exposed, it is our position that they be coated to achieve a fire resistant rating ecuivalent to the ceiling bcrrier.
021-18 021.29 You have nc t responded to the Appendix A guideline concerning fire doors 'being locked and alarmed. Confirm that you meet this pos i tion.
In particular, all fire doors used to protect openings in the wall separating the control building from the turbine building be alarmed and annunciated in the control room. These circuits should be electrically sucervised.
021.30 Cetermine whether the Jollapse of the turbine building rcof will affect the operaticn of any safety related equipment, particularly equiement in the auxiliary building adjacent to the wall between the auxiliary and turbine buildings.
021. 31 Provide a detailed description of the C0 system design criteria 2
and operation including concentrations, soaking times, and allowance for C0 leakage for each area where used. Also provide information 2
to demonstrate the CO system can protect against the potential 2
fires (cable fires, flammable liquid fires, or other fire hazards) identified within each area. For these areas, describe the provi-sion for personnel access following the C0 system discharge to 2
permit manual firefighting and cleanup if necessary.
e
+
021-19 ENCLOSURE 1 Mini. cn safe shutdewn systems when ene divisten of ali sa".w 1.0 s'"7it e-s is act avv i aci e:
1.1 Follcwing any fire, the plant can be brought to het shutdcwn conditiens using equip.ent and systams natars free of ffre damage.
1.2 The plant should be cacable of maintaining het shutdewn conditions fer an extanced time ;ericct.significantly longer than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
1.3 Fire damage to systems necessary to achieve and maintain cold shu::cwn conditions snculd be limited sc :na: repairs can be made and colc snut:cwn conditions acnievec witnin 72 hcurs.
1.4 Repair precedures for cold shutdcwn systems should be prepared new and material needed for such recairs should be en :ne site.
1.5 The het shu:dcwn condition must be achievable with pcwer frem the offsita ;cher systam, and ucen its Icss, wita ;cwer frem the ensita ;cwer system. A dedicated ;cwer su; ply may be substituted for :ne ensite ;cwer systs.
1.6 The ;cwer needed to achieve the ccid shutdcwn ecndition may be ce sined frem any one of the offsita pcwer, ensita pcwer, and dedicated ;cwer system.
1.7
'4 hen these ninimum systems art previced-their adecuacy shall be verified by 1 norcugn evaluaticn of:
a.
Systams recuired for het shu:dewn; b.
Systems escuired for caid shutdcwn; c.
Fire camage u ;cwer cistribu;icn systams; and d.
Interacticns caused by fire damage to ;cwer and water succly systams. anc.c succcr ing systes, i.e., cc=cenent ecciing
- atar succly.
- 2. 0 Mini um #dre c etacticn -nen dedicated or af ternate shutdcwn svsums are :rev' cec:
2.1 The fire pretacticn systams in artes (such as cable screacing recms) :na: cenuin camles for a large numcer cf systems snculd censis of:
a.
i n detecticn systam; b.
Hose s:s:1cns; and Fixec manual succetssien systam (gas or watar) c.
.IC~I : Ccnsiceration c :rsventing fire :recagatien-via c:verec trays, fin retartant ccating, tar-iers or ciannets en a casa-cy-casa basis.
212-21 Revised January 22, 1979 212.0 Reactor Systems Branch (ISB) 4 If this transient gives the limiting change in MCPR, the assumed maximum feedwater flow rate should be verified in the preoperational test program or the analysis should demonstrate that such a test is not needed to assure the safety limit will not be exceeded.
212.54 This question intentionally deleted.
(15.0) 212.52 Figure 15B.0-1 is not consistent with the text of section (15.0) 15B.0.3.3.3.
Provide the correct fioure and modifications to the text as required.
212.56 Provide results of an analysis to demonstrate that no sinole (6.3) failure will result in overpressurization of the RHR system.
Provide the design basis used to determine the capacity of the relief valves of the RHR system.
212.57 In section 5.5 of the FSAR, it is stated that the RCIC system (5.5) is designed to meet seismic Category 1 requirements. However, the suction of the RCIC pump is normally aligned to the condensate storage tank which does not meet seismic Category 1 requirements.
The HPCI system suction, which is also normally aligned to the condensate storage tank, is automatically transferred to the suppression pool if a low level in the condensate storage tank is reached. However, realignment of the RCIC suction requires manual action.
In the event of an SSE with concurrent loss of offsite power, sufficient time may not be available to permit credit for such manual actions to prevent unacceptable consequences.
Provide an analysis of the consequences of an SSE with concurrent loss of offsite power to demonstrate that the use of manual actions af ter 10 minutes to accomplish switchover of the RCIC suction line is acceptable or else modify the system by providing an acceptable water source (seismic Category ik or automatic transfer to the suopression pool. The analysis should be based i
e
..rfJF A 81 oYr"run am a..rr.r m n e s.. ma ' s. 7 mr.*ze.: w am:4 t.rn a r, ;r. s:. e u =..r r.;.. mza.a rwd r.wr.my GENERAL ATOMIC COMPANY P.O. BOX 81608 SAN DIEGO. CALIFORNIA 92138 (714) 455-3000 February 6, 1979 Mr. William Gammill Assistant Director for Advanc.ed Reactors Division of Project Management U. S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Mr. Gammill,
General Atomic Company is enclosing fifty (50) copies of its response to your Request for Additional Information (RAI) on H-451 Graphite, dated January 10, 1979.
No response to item 231.18 is being provided since it was presented in the RAI as a point of clarification only.
If you have any questions regarding these responses, please do not hesitate to let us know.
Sincerely, h
G. L. Wessman, Director Plant Licensing Division GLW:mk 7s0214.o0q3 e,4
STATE OF CALIFORNIA ss.
COUNTY OF SAN DIEGO After being duly sworn, the person known to me to be G. L. Wessman of General Atomic Company, signed the within document this 6 day of February 1979.
WITNESS my hand and official seal.
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RESPONSE TO NRC QUESTION 130.2 ON H-451 GRAPHITE QUESTION 130.2 In your response to Q130.la, you indicated that unirradiated H-451 graphite does not experience fatigue damage under reversed stress cycling unless stressed above a homologous stress limit of 0.63 in the axial direction or 0.74 in the radial orientation.
These homologous stress limits are understood to be established on the basis of 50 percent survival.
Indicate your rationale for not using limits for 99 percent survival.
If the limits for 99 percent survival are used, will your conclusion be still the same?
RESPONSE
The homologous fatigue stress limits quoted in the response to Question 130.la were inadvertently taken from the wrong section of Refer-ence 1.
The amended values of the H-451 graphite homologous fatigue stress limits for reversed stress cycling for 50% specimen survival at 105 cycles are 0.60 in the axial directiJn and 0.75 in the radial direc-tion (Ref.1).
The homologous fatigue stress limits for reversed stress cycling for 99% survival with 95% confidence at 105 cycles are 0.44 for the axial direction and 0.50 in the radial direction (Ref.1).
If these limits for 99% survival are used to evaluate the fatigue lives of the H-451 elements, the conclusions in the prior response to Question 130.la remain the same.
Both secondary and primary stresses have been calculated to be well below the stress limits for 99% survival, and during normal oper-ation no fatigue inducing mechanism exists.
REFERENCE 1.
Price, R.
J., " Cyclic Fatigue of Near Isotropic Graphite:
Influence of Stress Cycle and Neutron Irradiation", GA-A14588, November 1977.
RESPONSE TO NRC QUESTION 231.17 ON H-451 GRAPHITE QUESTION 231.17 The response to ist round question 231.3b indicated that because of the high degree of complexity of strain behavior in the radial orientation (as compared with axial strain behavior), no general comparison of stresses caused by irradiation strain in H-327 ar.d H-451 graphites has been provided.
Yet, the effects of these radial shrinkages are said to have been included in the analyses.
Notwithstanding the asserted complexity in radial strain behavior, some linkage must be provided, at least in the form of an exemplar calculation, between the radial strains and the stress provided in Table 2-1 of GLP-5588.
RESPONSE
The comparison of stresses caused by irradiation strains in H-327 and H-451 graphite is included in Table 2-1 of GLP-5588 for strains in both the axial and radial direction. The behaviors of the two graphPes in the axial direction were compared in the text of Section 2.4.1.4 because the irradiation strains are so well behaved in the axial direction that they can be described and compared in simple terms.
(The axial strain behaviors of H-327 and H-451 graphites are shown in Figures 4-1 and 4-5 (a) of GLP-5588).,
In contrast, the radial irradiation strain behaviors of both H-327 and H-451 graphite are complex (Figures 4-2 and 4-5 (b) of GLP-5588),
and it is difficult to provide a general description of this behavior in words. Although the radial strain behavior is not described in the text of Section 2.4.1.4 (GLP-5588), it is included in the calculation of stresses and is reflected in the analytical results.
All material properties, in-cluding the radial strain behavior, are included as input to the stress analysis computer codes, FESIC and SAFE GRAFITE. A detailed understanding of the impact of the radial strain behavior for the specific temperature and fluence history that produced the stresses listed in Table 2-1 can be obtained only by examining the details of the analysis.
RESPONSE TO NRC OVESTION 231.19 ON H-451 GRAPHITE QUESTION 231.19 (a) Although there are no data for the diffusion of strontium in H-451 graphite, it is stated in the report that, based on the data shown in Table 7 and Figure 3, "there appears to be no dependence on the type of graphite." Please indicate which of the referenced data are for near-isotropic graphites.
If little or no strontium diffusion data exist for near-isotropic needle-coke graphites, discuss the applicability of the data to H-451 graphite.
(b) The strontium diffusion data in Figure 3 of the report have a fairly wide scatter (1 to 2 orders of magnitude) in the lower range of temperatures whereas at high temperatures the scatter is smaller.
Please indicate the range of expected operating temperatures in the Fort St.
Vrain H-451 blocks.
RESPONSE TO 231.19(a)
None of the data shown in the response to Question 231.12 are for the dif-fusion of strontium in near-isotropic graphite.
The conclusion that strentium diffusion in H-451 graphite does not differ significantly from that in H-327 graphite is inferred based on the fact that the available data, taken on many graphites, indicate no dependence of strontium diffusivity upon graphite type.
RESPONSE TO 231.19(b)
As noted in the response to Question 231.16c, the expected range of operat-ing temperatures in the Fort St. Vrain H-451 fuel blocks is 420-1210 C.
It should be noted that most of the scatter in strontium diffusivity data at low temperatures i: below the fit to the data shown in Figure 3 of the response to Question 231.12.
RESPONSE TO NRC QUESTION 231.20 ON H-451 GRAPHITE QUESTICN 231.20 (a)
The report indicates that, whereas strontium diffusion in graphite can be described by Fick's Law, cesium transport is a more complex phenomenon.
Yet cesium transport is described in the report in terms of a permeation coefficient, defined by the equation J = n (C1-C )/L, 2
(report equation 9), which is a fonn of Fick's Law.
Please discuss this apparent contradiction; that is, discuss how a process that is acknowledged to be more complex, can be described by Fick's Law, when Fick's Law is known to be generally applicable only to very simple diffusion case:
(b) There appears to be only one datum (see report Figure 4) indi-cating the effect of irradiation on cesium permeation coefficients for H-451 graphite and none for H-327.
It is difficult, therefore, to accept the conclusion that "the permeation coefficient for Cs in H-451 graphite is significantly reduced in-pile." Please discuss the effect of using the permeating coefficients for unirradiated graphite as interim values until more data are obtained on cesium diffusion in H-451 in-pile.
Please indicate whether such data are to be ob-tained on the 8 test elements currently scheduled for the first reload in Fort St. Vrain.
RESPONSE TO 231.20(a)
The definition of the permeation coefficient, AC _ n(C1-Cp)
II)
J=
-n g -
L is analogous but not equivalent to the definition of the diffusion coefficient where Fick's laws apply.
In Eq. (1), J is the. flux, - the permeation coeffi-cient, C1 the concentration of the diffusant at the source boundary of the medium in which diffusion occurs, C2 the concentratier, at the sink boundary, and '. the thickness of 'he medium, assuming slab geometry.
Fick's first law of diffusion is (Ref.1)
J=
-0(3C/aX).
(2)
Consider, for simplicity, the steady-state condition for a slab of graphite of thickness L; also, let C2 = 0.
Then, Fick's first law becomes J = DC1/L.
(3)
Fick's laws (first and second) also imply that the loading of the graphite per unit area is M = C1L/2 (4) where M is the loading.
Thus, the flux and loading are related according to 2
J = 2DM/L.
(5)
It is this combination of flux and leading which is characteristic of diffu-sion in media '.o which Fick's laws apply under steady-state conditions. Anal-ogous relath..,s can be obtained for non-steady-state conditions. When Fick's Laws apply, steady-state loading and mass flux are simultaneously attained.
The permeation coefficient (Eq. 1) can be used to describe the rate of Cs migration through and release from graphite. However, nothing is implied by Eq. 1 about the loading of the graphite.
The fact that Cs loadings observed when studying Cs diffusion in graphite (Ref. 2) did not attain steady-state values at the same time as the Cs mass flux at the sink boundary reached its steady-state value leads to the conclusion that Cs diffusion does not obey Fick's laws.
RESPONSE TO 231.20(b)
Transport data for cesium in H-327 graphite in-pile are available (Refs.
2-4) and also support the conclusion that cesium transport is significantly reduced in-pile.
In the case of H-451 graphite, the one comparison between in-pile and out-of-pile transport was based on two experiments in which the same apparatus and source loading were used in-pile and out-of-pile.
Nevertheless, the effect of using permeation coefficients for unirradiated H-451 graphite as interim values has been examined. As shown in Figure 4 of the response to Question 231.12, the permeation coefficient of cesium in unir-radiated H-451 graphite (solid line) is greater than values given in the FSAR (broken line) by about a factor of seven. The cesium source terms in the FSAR are based on the permeation coefficients given in the FSAR.
If it is assumed,
as a first approximation, that cesium release is directly proportional to the permeation coefficient in graphite, then the calculated cesium-137 " Design" plateout inventory of 5460 Ci given in Table 3.7-2 of the FSAR becomes approxi-mately 38,000 Cf.
The effect of this increased cesium source term on accident consequences discussed in Chapter 14 of the FSAR was investigated. The two-hour exclusion area boundary (EAB) bone dose for Design Basis Accident No. 2, Rapid Depres-surization/Blowcown, is the only dose significantly affected by the hypothctical increase in cesium inventory.
Using the same calculational methods used in the FSAR analysis, the two-hour EAB bone dose for DBA-2 is determined to be increased by about 0.5 rem if the Cs-137 inventory is increased to about 38,000 Ci.
The increase is relatively small because the bone dose effectiveness of Cs-137 compared to strontium-90 is relatively small.
The resultant bone dose remains will below accepted limits.
With regard to test elements FTE-1 through FTE-8, the currently planned DOE-funded post-irradiation examination program (PIE), described in Amendment 2 (Appendix A) of General Atomic Report GLP-5494, includes gamma spectroscopy of graphite components and destructive fission product examinations.
These tests will allow the general characterization of fission product migration, including cesium transport through H-451 graphite.
Although analysis of data obtained from the test elements can be used to check our ability to calculate fission product migration in HTGR fuel within the uncertainties of PIE measurements, material properties, and operating con-ditions, this analysis will probably not enable accurate determination of permeation parameters.
Precise determination of permeation coefficients requires accurate control of test conditions, including temperature histories and fission product source concentratlons.
Such control is not attainable during test element irradiation in the FSV core.
REFERENCES 1.
J. Crank, "The fiathematics of Diffusion", Oxford University Press,1956.
2.
D. Chandra and J. H. Norman, " Diffusion of Cesium Through Graphite",
J. Nucl. tiater j2, 293-310 (1976).
3.
C. F. Wallroth, et.al., " Post Irradiation Examination of Peach Bottom Fuel Test Element FTE-3", USAEC Report GA-A13004, General Atomic Company, August 15, 1974.
4.
C. F. Wallroth, et.al., " Post Irradiation Examination of Peach Bottom Fuel Test Element FTE-6", USERDA Report GA-A13943, General Atomic Company, September 1977.
RESPONSE TO NRC QUESTION 231.21 ON H-451 GRAPHITE QUESTION 231.21 The response to 1st-round Question 231.16 indicates (Table 1) that the creep tests on H-451 graphite will not be completed until 1983, whereas H-451 graphite reload fuel elements could be placed in Fort St. Vrain as soon as late 1980 or early 1981.
Moreover, much of the current creep data base on near-isotropic graphite appears to have been generated on Gilsocarbon-based, near-isotropic graphite rathar than petroleum coke-based, near-isotropic graphites.
In view of (a) the current limitations on the data base on H-451 graphite and (b) the uncertainty regarding the applicability of Gilsocarbon graphite creep data, please indicate how the test data yet to be developed would lead the expected irradiation exposures of reload elements.
Also, show how the planned post-irradiation examination program on the 8 test elements and future reload elements will provide dimensional change data that would be used to verify the pre-dicted changes that are based, in part, on the expected creep behavior.
RESPONSE
Based on current projections, earliest possible use of H-451 graphite in Fort St. Vrain reload segment 9 would begin in early 1981.
Figure 1 shows the schedule for completion of H-451 graphite irradiation creep tests at JRC Petten and at ORNL.
Creep tests to 7.4 x 1021n/cm2 (E >
0.18 MeV, HTGR) at 840*C in the High Flux Reactor, Petten, will be com-pleted at about the end of 1979 and confirmatory tests at 900 C at ORNL will reach full fluence in mid-1981.
(The latter test was inadvertently omitted from the response to Question 231.16.) 600*C creep test data 2
from ORNL will be available to 3 x 102I n/cm by mid-1980 and full exposure data at 600 C will be available in mid-1982.
ORNL creep tests at 1250 C, planned for a later date, are not directly re!nant to FSV operation since most of the graphite in FSV operates at significantly lower tempcratures. Thus, full exposure creep data at 840-900 C will be available in time for the earliest use of H-451 in Fort St. Vrain.
Full exposure creep data at 600 C will be available when the first H-451 reload segment has accumulated about 1 x 1021 n/cm2,
The currently planned DOE-funded post-irradiation examination (PIE) program for test elements FTE-1 through FTE-8 is described in Amendment 2 (Appendix A) of General Atomic Report GLP-5494, June 30,1977. The pro-gram includes detailed graphite block metrology for all eight test elements.
Data will be recorded on magnetic tape and will be compared with pre-irradiation measurements to establish irradiation induced graphite strain and bow.
Measured strain and bow can be compared with strain and bow calcu-lated for each test element to provide an indication of the validity of such calculations. The measured strain and bow should agree with calcula-tions within the uncertainties of the PIE measurements and uncertainties in material properties and operating conditions.
Input to such calculations, however, includes numerous physicci properties of graphite; it is, accord-ingly, not possible to assess the accuracy of any single input property, such as creep behavior, by examining behavior which is simultaneously dependent upon several properties.
The irradiation creep tests described above and in the response to question 231.16 will provide the primary data base for H-451 graphite creep behavior.
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