ML19274F082
| ML19274F082 | |
| Person / Time | |
|---|---|
| Issue date: | 05/09/1979 |
| From: | Arlotto G NRC OFFICE OF STANDARDS DEVELOPMENT |
| To: | Fraley R Advisory Committee on Reactor Safeguards |
| Shared Package | |
| ML19274F083 | List: |
| References | |
| RULE-PR-50 NUDOCS 7906130075 | |
| Download: ML19274F082 (15) | |
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- 379 MEMORANDUM FOR
Raymond F. Fraley, Executive Director, ACRS FROM:
Guy A. Arlotto, Director, DES, SD
SUBJECT:
PROPOSED GENERAL RFVISION OF APPENDIX G, " FRACTURE TOUGHNESS REQUIREMENTS" AND APPENDIX H, " REACTOR VESSEL SURVEILLANCE PROGRAM REQUIREMENTS" Enclosed for the use of the Subcommittee on Regulatory Activities are 15 copies of the draft amendments, which constitute a general revisica of Appendices G and H to 10 CFR Part 50.
In October,1978, the Subcommittee reviewed a " limited revision" of Appendices G and H, which modified the fracture toughness requirements for bolts and lifted restrictions on the location and method of attach-ment of surveillance capsule holders. The public comment period on those proposed changes ends on May 14, 1979.
Meanwhile, the enclosed' general revision has been prepared to meet several long-standing needs.
In the use of Appendices G and H since they became effective in August, 1973, a number of the requirements have been found to need clarifications in language. Six years of use has also shown that certain restrictions in Appendices 6 and H can and should be modified.
Finally, there have been ghanges in the ASME Code and in ASTM E 185 that need to be reflected it/ Appendic G and H.
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I' Guy A.
- rlotto, rector Divisio of Engineering Standards Office of Standards Development
Enclosures:
ySiicDocument com 7 W. S. Hazelton, D0R 1.
Notice of Rule Making H. R. Denton, DSE W. J. Collins, IE 2.
Marked-up copy of Appendices H. D. Thornburg, DRCI G and H N. C. Moseley, Director, DROI 3.
Value/ Impact Statement J. M. Felton, Director, 0AM:DRR R. M. Gamble, DSS
Contact:
P. N. Randall S. Levine, RES 443-5997 V. Stello, Jr., D0R R. S. Boyd, DPM 790613007f 1
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NUCLEAR REGULATORY COMMISSION (10 CFR Part 50)
DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Fracture Toughness Requirements for Nuclear Power Reactors AGENCY:
U.S. Nuclear Regulatory Commission ACTION:
Proposed rule
SUMMARY
The Nuclear Regulatory Commission (NRC) is considering amend-ing its regulations which specify fracture toughness requirements for nuclear power reactors and its requirements for reactor vessel mate-rial surveillance programs. The amendments would clarify the applica-bility of these requirements to old and new plants, modify some require-ments that have proved to be unduly conservative, and shorten and simplify these regulations by more extensively incorporating by refer-ence appropriate National Standards.
DATES:
Comment period expires 1979.
ADDRESSES: Written comments should be submitted to the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, D.C.
20555, Attention:
Docketing and Service Branch.
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FOR FURTHER INFORMATION CONTACT:
Dr. P. N. Randall, Office of Stand-ards Development, U.S. Nuclear Regulatory Commission, Washington, D.C.
20555, 301-443-5997.
SUPPLEMENTARY INFORMATION: When 10 CFR Part 50 was amended by adding Appendix G, " Fracture Toughness Requirements," and Appendix H, " Reactor Vessel Material Surveillance Program Requirements," on July 17, 1973 (38 FR 19012), the ASME Boiler and Pressure Vessel Code (the ASME Code) provisions pertaining to nuclear power plant components were being extensively revised.
Some requirements that had not yet become effec-tive in the ASME Code were included in Appendix G.
Now that the NRC has determined that the ASME Code requirements adequately reflect the NRC position, Appendix G is being condensed by more extensively incor-porating the ASME Code by reference.
Appendix H incorporated by reference the 1973 edition of ASTM E 185.1 Now there is a new edition, E 185-79. As amended, Appendix H would specify the earliest edition of E 185 that could be used for each part of the surveillance program, and would delete paragraphs that are now covered by specific provisions of E 185-79.
Paragraph 50.55a(i), which makes the provisions of Appendices G and H conditions of a construction permit for a utilization facility (e.g., a nuclear power plant), would be deleted and a new 6 50.47 IStandard Recommended Practice for Surveillance Tests for Light Water Cooled Nuclear Power Reactor Vessels.
Copies may be obtained from the American Society for Testing and Materials, 1916 Race St., Philadelphia, Pa. 19103.
Copies will be available for inspection at the Commission's Public Document Room, 1717 H St., NW., Washington, D.C.
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i would be added for this purpose.
This change would avoid any confusion concerning the applicability of Appendices G and H that might occur when S 50.55a is amended for purposes not relevant to the appendices.
Significant differences between the existing regulations in Appendix G and the proposed amendments are:1 SII.G.
The definition of " adjusted reference temperature" would be changed to use the 30 ft lb level instead of the 50 ft lb level on the curve of absorbed energy versus temperature.
When the upper shelf energy level approaches 50 ft ib, as it has in some test data, the definition based on 50 ft lb becomes invalid.
(See also the discussion of paragraph IV.B.)
SII.H.
The definition of " beltline" would be changed to avoid unnecessary materials tests of portions of the vessel above and below the core.
II.J.
The definition of " integrated surveillance program" would be removed from Appendix G and a more complete discussion of such a program would be incorporated in paragraph II.C.
of Appendix H.
SIII.
Section III, " Fracture Toughness Tests," would be reduced in length because most of the pertinent requirements could be incorporated by referencing the ASME Code.
HIII.A.
Language would be added to paragraph III. A. to clarify how the fracture toughness test requirements would be applied to "old" plants, those for which the reactor vessel was constructed to Listed according to existing paragraph numbers.
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's an ASME Code earlier than the Summer 1972 Addenda to the 1971 Edition.
This language would clarify the Commission's intention that all plants must meet the required margins of safety and the other fracture prevention requirements of Appendices G and H; but the owners of "old" plants may, when approved by the Commission, use fracture toughness tests and fracture analyses other than those required by Appen-dices G and H to demonstrate that the required margins of safety have been met.
Although, the present language of Appendices G and H does not so provide, the statement of considerations accompanying the adoption of these Appen-dices on July 17, 1973, (38 FR 19013) indicated that the Commission intended the appendices to have this flexibility.
Licensing actions since that time consistently reflected this policy until recently when legal questions arose con-cerning the need for exemptions to certain provisions of Appendices G and H for plants constructed to an ASME Code edition and addenda earlier than the Summer 1972 Addenda to the 1971 Edition.
It is to avoid the need for exemptions on this issue that the revision of paragraph III.A. is proposed.
SIV.
As amended,Section IV, " Fracture Toughness Requirements" would incorporate by reference the fracture toughness require-ments of the ASME Code and then list the requirements of this appendix as supplemental requirements, thereby deleting much technical detail from the regulation.
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'e VIV.A.2.c.
For boiling water reactors, the criticality limit based on the minimum permissible hydrotest temperature would be replaced by a lower limit to avoid unnecessary delay in startup of boiling water reactors.
HIV.A.2.d.
The minimum permissible temperature for system hydrostatic pressure tests would be reduced when performed when there is no fuel in the reactor to improve the efficiency of the inspection operation.
11V.B.
The 50 ft lb requirement for Charpy upper-shelf energy would be stated explicitly.
Language would be added to permit acceptance of a lower value without the need for an exemption if a fracture analysis showed that the margin of safety against ductile fracture was comparable with the margin of safety against fracture in the transition region required by para-graph IV.A.2.
SV.B.
A requirement would be added that material toughness values to be used in fracture analyses are those predicted for the material near the tip of the assumed flaw at its deepest part.
Significant differences between the existing regulations in Appendix H and the proposed amendments are:
SII.B.
As revised, paragraph II.B. would incorporate by reference ASTM E 185, and reference to the 1973 edition of ASTM E 185 would be deleted.
language would be added to permit the use of the 1979 edition or an earlier edition provided it was the current edition at the time the action was taken.
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For example, the earliest edition of ASTM E 185 that could be used in the selection of surveillance materials, preparation of specimens, and construction of surveillance capsules would be the edition that was current on the issue date of the ASME Code to which the reactor vessel was purchased.
UII.C.3.
Most of this paragraph would be deleted, because the requirements for withdrawal schedules contained in the 1979 editic, of ASTM E 185 are satisfactory.
TII.C.4.
The existing provision permitting integrated surveillance programs would be broadened by adding general requirements and criteria.
SIII.
Section III would be deleted because its provisions are covered in Appendix G,Section V.
,SIV.
Reporting requirements for the test results from each cap-sule withdrawal would be clarified and a time limit for sub-mittal of the report would be specified.
The Commission's Office of Standards Development has prepared a value/ impact statement for the proposed amendment, which provides additional technical details and justification.
This statement is available for inspection by the public in the Commission's Public Document Room at 1717 H Street, NW., Washington, D.C.
Single copies of the value/ impact statement may be obtained by request addressed to the Office of Standards Development, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, Attention:
P. N. Randall.
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Pursuant to the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and section 553 of Title 5 of the United States C)de, notice is hereby given that adoption of the following amendments of 10 CFR Part 50 is contemplated. All interested persons who wish to submit written comments o, suggestions in connection with the proposed amendments should send them to the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, Attention:
Docketing and Service Section by 1979.
Copies of comments received may be examined in the Commission's Public Document Room at 1717 H Street, NW., Washington, D.C 1.
Section 50.55a is amended by deleting paragrcph (i).
2.
A new Section 50.47 is added to 10 CFR Part 50 to read as follows:
9 50.47 - Acceptance Criteria for Fracture Prevention Measures for Light Water Nuclear Power Reactors (a) (1)
Except as provided in paragraph (a)(2) of this section all light-water nuclear power reactors shall meet the fracture toughness and material surveillance program requirements for the reactor coolant pressure boundary set forth in Appendices G and H to this part.
(2)
Proposed alternatives to the described requirements or portions thereof may be used when authorized by the Commission upon demonstration by the applicant that (i) compliance with the specified requirements would result in hardships or unusual difficulties without a compensating increase in the level of quality and 7
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safety and (ii) the proposed alternatives would pro-vide an acceptable level of quality and safety.
3.
Appendices G and H to 10 CFR Part 50 are amended to read as follows:
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APPENDIX G FRACTURE TOUGHNESS REQUIREMENTS I.
INTRODUCTION AND SCOPE This appendix specifies minimum fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary of light water nuclear power reactors to provide adequate margins of safety during any condition of normal ope-ration, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime.
The requirements of this appendix apply to the following materials:
A.
Carbon and low-alloy ferritic steel plate, forgings, castings, and pipe with specified minimum yield strengths not over 50,000 psi (345 MPa), and to those with specified minimum yield strengths greater than 50,000 psi (345 MPa) but not over 90,000 osi (621 MPa) if quali-fied by using methods equivalent to those described in paragraph G-2110 of the ASME Code (Defined in paragraph II.A. below)1.
B.
Welds and weld heat-affected zones in the materials specified in section I.A.
C.
Materials for bolting and other types of fasteners with specified minimum yield strengths not over 130,000 psi (896 MPa).
1The latest Edition and Addenda permitted by paragraph 50.55a(b) at the time the analysis is made shall be used for the purposes of paragraph I.A.
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Adequacy of the fracture toughness of other ferritii: materials shall be demonstrated to the Commission on an individual case basis.
II.
DEFINITI0MS A.
"ASME Code" means the American Society of Mechanical Engineers Boiler and Pressure Vessel Code.
If no section is specified, the reference is to Sectiot, III, Division 1, " Rules for Construction of Nuclear Power Plant Components." "Section XI" meansSection XI, Division 1, " Rules for Inservice Inspection of Nuclear Power Plant Components." If no edition or addenda is specified, the applicable l
ASME Code edition and addenda and any limitations and modifications thereof are specified by S 50.55a, Codes and Standards.
B.
"Ferriticmaterial"meanscarbonandlow-alioysteels, higher alloy steels including all stainless alloys of the 4xx series, and maraging and precipitation hardening steels with a predominantly body-centered cubic structure.
C.
" System hydrostatic tests" means all preoperational system leakage and hydrostatic pressure tests and all system leakage and hydrostatic pressure tests performed during the service life of the pressure boundary in compliance with the ASME Code,Section XI.
1Copies may be obtained from the American Society of Mechanical Engineers, United Engineering Center, 345 East 47th Street, New York, N.Y. 10017.
Copies are available for inspection at the Commission's Public Document Room, 1717 H St. N.W., Washington, D.C.
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D.
"Specified minimum yield strength" means the minimum yield strength (in the unirradiated condition) of a material specified in the construction code under which the component is built pursuant to 9 50.55a.
E.
" Reference temperature" means the reference temperature, RTNDT, as defined in the ASME Code.
F.
" Adjusted reference temperature" means the reference temperature as adjusted for irradiation effects (see Section V of this Appendix) by adding to RT the temperature shift in the average NDT Charpy curve for the irradiated material relative to that for the unirradiated material, measured at the 30 ft lb (41J) level.
G.
" Beltline" or " Beltline region of reactor vessel" means the region of the reactor vessel (shell material including weld regions and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions that are predicted to experi-ence sufficient neutron radiation damage to be considered in the selec-ti'en of the most limiting material with regard to radiation damage.
III.
FRACTURE TOUGHNESS TESTS A.
To demonstrate compliance with the minimum fracture toughness requirements of Sections IV and V of this appendix, ferritic materials shall be tested in accordance with the ASME Code and, for the beltline materials, the test requirements of Appendix H.
For a reactor vessel that was constructed to an ASME Code earlier than the Summer 1972 11 g7;.I
Addenda of the 1971 Edition (pursuant to 650.55a), the program of frac-ture toughness tests and analyses shall be supplemented in a manner approved by the Commission to demonstrate compliance with the fracture toughness requirements of this appendix.
B.
Test methods for supplemental fracture toughness tests des-cribed in paragraph V.C.2. shall be submitted to and approved by the Commission prior to testing.
C.
All fracture toughness test programs conducted in accordance with paragraphs A and B of this Section shall comply with ASME Code requirements for calibration of test equipment, qualification of test personnel, and retention of records of these functions and of the test data.
IV.
FRA'CTURE TOUGHNESS REQUIREMENTS A.
The pressure-retaining components of the reactor coolant pres-sure boundary that are made of ferritic materials shall meet the frac-ture toughness requirements of the ASME Code supplemented as follows for fracture toughness during system hydrostatic tests and any condi-tion of normal operation, including anticipated operational occurrences.
1.
Reactor vessel beltline materials shall have minimum Charpy upper-shelf energy of 75 ft lb (102J) initially and shall main-tain minimum upper-shelf energy throughout the life of the vessel of 50 ft lb (68J), unless it is demonstrated in a manner approved by the 12 5:. '.
Commission that lower values of upper-shelf energy will provide margins of safety against fracture comparable with those required by paragraph 2.
2.
When the core is not critical, pressure-temperature limits for the reactor vessel shall be at least as conservative as those obtained by following the methods of analysis and the required l
margins of safety of Appendix G of the ASME Code supplemanted by the requirements of Section V of this Appendix.
The justification sub-mitted for the pressure temperature limits shall describe the method of analysis used and shall demonstrate that when the limits are con-trolled by the nozzles, flanges and shell regions near structural discontinuities, the neargins of safety for those regions are comparable with those required for the beltline when it is controlling.
3.
When the core is critical (other than for the purpose of low-level physics tests) the temperature of the reactor vessel shall not be lower than 40 F above (22 C above) the minimu:a permissible temper-ature of paragraph 2. of this section nor lower than (i) for pressurized water reactors, the minimum permissible temperature for the inservice system hydrostatic pressure test, or (ii) for boiling water reactors when water level is within the normal range for power operation, 60 F above (33 C above) the reference temperature of the closure flange regions that are highly stressed by the bolt preload.
4.
If there is no fuel in the reactor during system hydro-static pressure tests or leak tests, the minimum permissible test temper-ature shall be RTriDT + 60 F (RTtIDT + 33 C).
1The latest Edition W Addenda permitted by paragraph 50.55a(b) at the time the analysis is made chall be used for the purposes of paragraph IV.A.2.
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5.
If there is fuel in the reactor during system pressure tests or leak tests, the requirements of paragraphs 2 or 3 shall apply, depending on whether the core is critical during the test.
B.
Reactor vessels for which the predi:ted value of upper shelf energy at end of life is below 50 ft lbs or the predicted value of adjusted reference temperature at end of life exceeds 200 F (93 C) shall be designed to permit a thermal annealing treatment to recover material toughness properties of ferritic materials of the reactor vessel beltline.
V.
INSERVICE REQUIREMENTS - REACTOR VESSEL BELTLIflE MATERIAL A.
The effects of neutron radiation on the reference tempera-ture and upper shelf energy of reactor vessel beltline materials, including welds, shall be predicted from the results of pertinent radiation effects studies in addition to the results of the surveil-lance program of Appendix H.
8.
Reactor vessels may continue to be operated only for that service period within which the requirements of Section IV are satis-fled using the predicted value of the adjusted reference temperature and the predicted value of the upper-shelf energy at the end of the service period to accourt for the effects of radiation on the fracture toughness of the beltline materials.
These predictions shall be made for the radiation conditions at the tip of the assumed flaw at its 14
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deepest part.
The highest adjusted reference temperature and the lowest upper-shelf energy level of all the beltline materials shall be used to verify that the fracture toughness requirements are satisfied.
C.
In the event that the requirements of Section V.B. cannot be satisfied, reactor vesseis may continue to be operated provided all of the following requirements are satisfied:
1.
Volumetric examination of the beltline materials that do not satisfy the requirements of Section V.B. including 100 percent of any associated welds shall be made and any flaws evaluated accord-
'ing to Section XI of the ASME Code and as specified by the Commission.
2.
Additional evidence of the fracture toughness of the beltline materials after exposure to neutron irradiation shall be obtained from results of supplemental fracture toughness tests.
3.
An analysis shall be performed that conservatively demonstrates, making appropriate allowances for all uncertainties, the existence of adequate margins of safety for continued operation.
D.
If the procedures of section V.C. do not indicate the exis-tence of an adequate safety margin, the reactor vessel beltline may be subjected to a thermal annealing treatment to recover the fracture toughness of the material.
The degree of such recovery shall be mea-sured by testing additional specimens that have been withdrawn from the surveillance program capsules and have been annealed under the same time-at-temperature conditions as those given the beltline mate-rial.
The results shall provide the basis for establishing the adjusted reference temperature after annealing.
The reactor vessel 15 M*
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may continue to be operated only for that service period within which the predicted fracture toughness of the beltline region materials satisfies the requirements of Section IV. A. using the values of adjusted reference temperature and upper-shelf energy that include the effects of annealing and subsequent irradiation.
E.
The proposed programs for satisfying the requirements of Sections V.C. and V.D. shall be reported to the Director of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, for review and approval on an individual case basis at least 3 years prior to the date when the predicted fracture toughness levels will no longer satisfy the requirements of Section V.B.
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APPENDIX H - REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM REQUIREMENTS I.
INTRODUCTION The purpose of the material surveillance program required by this appendix is to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region of light water nuclear power reactors resulting from their exposure to neutron irradia-tion and the thermal environment.
Under this program, fracture tough-ness test data are obtained from material specimens exposed in surveil-lance capsules, which are withdrawn periodically from the reactor vessel.
These data will be used as described in Sections IV and V of Appendix G.
II.
SURVEILLANCE PROGRAM CRITERIA A.
No material surveillance program is required for reactor vessels for which it can be conservatively demonstrated 'by analytical methods applied to experimental data and tests performed on comparable vessels, making appropriate allowances for all uncertainties in the measurements, that the peak neutron fluence (E>1MeV) at the end of the design life of the vessel will not exceed 1017 n/cm,2 B.
Reactor vessels that do not meet the conditions of paragraph II.A. shall have their beltline materials monitored by a 17
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1 surveillance program complying with ASTM E 185, except as modified by this Appendix.
1.
That part of the surveillance program conducted prior to the first capsule withdrawal shall meet the requirements of the edition of ASTM E 185 that is current on the issue date of the ASME Code to which the reactor vessel was purchased.
For each capsule withdrawal, the test procedures and reporting requirements shall meet the requirements of the edition of E 185 in effect on the date of capsule withdrawal, to the extent practical for the configuration of the specimens in the capsule.
For any part of the surveillance program, later editions of E 185 including only those editions through 1979 may be used instead of the editions specified above.
2.
Surveillance specimen capsules shall be located near the inside vessel wall in the beltline region, so that the specimen irradiation history duplicates to the extent practicable, within the physical constraints of the system, the neutron spectrum, temperature history, and maximum neutron fluence experienced by the reactor vessel inner surface.
If the capsule holders are attached to the vessel wall or to the vessel cladding, construction and inservice inspection of the attachments and attachment welds shall be done according to the 1Standard Recommended Practice for Surs eillance Tests for Light Water Cooled Nuclear Power Reactor Vessels.
Copies may be obtained from the American Society for Testing and Materials, 1916 Race St., Philadelphia, Pa. 19103.
Copies will be available for inspection at the Commission's Public Document Room, 1717 H dt., NW., Washington, D.C.
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requirements for permanent structural attachments to reactor vessels given in the ASME Code, Sections III and XI.
The design and location of the capsules shall permit insertion of replacement capsules.
Accelerated irradiation capsules may be used in addition to the required number of surveillance capsules specified in ASTM E 185.
3.
Proposed withdrawal schedules shall be submitted with a technical justification therefor, to the Commission for approval.
The proposed schedule shall not be implemented without prior Commis-sion approval.
C.
An integrated surveillance program may be considered for a set of reactors that have similar design and operating features.
The materials chosen for surveillance from each reactor in the set may be irradiated in one or more of the reactors.
No reduction in the requirements for number of materials to be irradiated, specimen types, or number of specimens is permitted, but the amount of testing may be reduced if the initial results agree with predictions.
Integrated surveillance programs must be approved by the Commission on a case-by-case basis.
Criteria for approval include the following considera-tions:
(1) the design and operating features of the reactors in the set must be sufficiently similar to permit accurate comparisons of the predicted amount of radiation damage as a function of total power out-put, (2) there must be adequate arrangements for data sharing between plants, and (3) there must be substantial advantages to be gained in the predicted reductions in power outages and personnel exposure to radiation.
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III.
REPORT OF TEST RESULTS A.
Each capsule withdrawal and the results of the fracture toughness tests shall be the subject of a summary technical report to be provided to the Director of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, D.C.
20555, within 90 days after completion of testing.
The Commission shall be notified at least 30 days in advance of the capsule withdrawal, giving the expected date of completion of testing and submittal of report for approval by the Commission.
B.
The report shall include the data required by ASTM E 185 and the results of all fracture toughness tests conducted oa the beltline materials in the irradiated and unirradiated conditions.
C.
If a change in the operating pressure-temperature limits given in the Technical Specifications is required, the revised limits shall be submitted with the report, including any changes made in operating procedures required to meet such limits.
(Secs. 103, 104, 1611, Pub. Law 83-703; 68 Stat. 936, 937, 948; Sec. 201, Pub. Law 93-438, 88 Stat. 1242; (42 U.S.C. 2133, 2134, 2201(i), 5841).)
Dated at this day of 1979.
For the Nuclear Regulatory Commission.
Samuel J. Chilk Secretary of the Commission 20
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