ML19290A329
| ML19290A329 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/14/1976 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19290A327 | List: |
| References | |
| NUDOCS 7911060571 | |
| Download: ML19290A329 (13) | |
Text
.
f UNITED STATES
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 2o666 METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER AND LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY DOCKET NO. 50-289
'lliREE MILE ISLAND NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 15 License No. DPR-50 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Metropolitan Edison Company, Jersey Central Power and Light Company, and Pennsylvania Electric Company (the licensees) dated March 23, 1976, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
An environmental statement or negative declaration need not be prepared in connection with the issuance of this amendment.
1555 249 1911060 f7/
. 2.
Accordingly, the license is amended by a change to the Technical Specifications as indicated in the attachment to this license amendment.
3.
This license amendment is effective as of the date of its issuance.
FOR 111E NUC R REGULATORY C0501ISSION Robert W. Reid, Chief Operating Reactors Branch No. 4 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance:
g 1555 250
ATTACHMENT TO LICENSE AMENDMENT NO.15 FACILITY OPERATING LICENSE NO. DPR-50 DOCKET NO. 50-289 Revise Appendix A as follows:
Remove pages 3-3, 3-4, 3-5, 4-11, 4-12, 4-13, and 4-14 and insert attached pages 3-3, 3-4, 3-5, 4-11, 4-12, 4-13 and 4-14.
Changes on the revised pages are shown by marginal lines. Pages 4-12 and 4-14 are unchanged and are included for convenience only.
1555 251
3.1.2 PRESSURIZATION, HEATUP, AND COOLDOWN LIMITATIONS Applicability Applies to pressurization, heatup, and cooldown of the reactor coolant system.
Objective To assure that temperature and pressure changes in the reactor coolant system do not cause cyclic loads in excess of design for reactor coolant system components.
Specification 6
3.1.2.1 For the first 1.7 x 10 thermal megawatt days (approximating two years) the reactor coolant pressure and the system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figure 3.1-1 and Figure 3.1-2 and are as follows:
Heatup:
Allowable combinations of pressure and temperature shall be to the right of and below the limit line in Figure 3.1-1.
Heatup rates shall not exceed those shown on Figure 3.1-1.
Cooldown:
Allowable combinations of pressure and temperature for a specific cooldown shall be to the left of and below the limit line in Figure 3.1-2.
Cooldown rates shall not exceed those shown on Figure 3.1-2.
_ Hydro Tests:
For isothermal system hydrotests during the first two years of operations, the system may be pressurizcd to the limits set forth in Specification 2.2, when there are fuel assemblies in the vessel and to ASME Code Section III limits when no fuel assemblies are present if the system temperature is 215 F or greater.
The cystem may be tested to a pressure of 1150 psig provided system temperature is 175 F or greater.
Initial system hydrotests prior to criticality may be conducted if the reactor coolant system temperature is 118 F or greater.
3.1.2.2 The secondary side of the steam generator shall not be pressuirzed above 200 psig if the temperature of the steam generator shell is below 100 F.
3.1.2.3 The pressurizer heatup and cooldown rates shall not exceed 1000F in any one hour.
The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 430F.
3.1.2.4 Within two effective full power years of operation, Figure 3.1-1 and 3.1-2 shall be updated in accordance with criteria acceptable to the NRC.
Bases All reactor coolant system components are designed to. withstand the effects of cyclic loads due to system temperature and pressure changes. (1)
These cyclic 1555 252 Amendment No. 15 3-3
loads are introduced by unit load transients, reactor trips, and unit heatup and cooldown operations.
The number of thermal and loading cycles used for design purposes are shown in Table 4-8 of the FSAR. The maximum unit heatup and cooldewn rate of 100 F in any one hour satisfies stress limits for cyclic operation.(2)
The 200 psig pressure limit for the secondary side of the steam generator at a 0
temperature less than 100 F satisfies stress levels for temperatures below the DTT.(3)
The reactor vessel plate material and welds have been tcsted to verify conformity to specified requirements and a maximum NDM value of 30 F has been determined based on Charpy V-notch tests. The maximum NDTT value obtained for the steam generator shell material and welds was 40 F.
The heatup and cooldown rate limits in this specification are not intended to limit instantaneous rates of temperature change, but are intended to limit tempera-ture changes such that there exists no one hour interval, in which a temperature change greater than the limit takes place.
Figures 3.1-1 and 3.1-2 contain the limiting reactor coolant system pressure-temperature relaticnship for operation at Dn (4) and below to assure that stress Icvels are low enough to preclude brittle fracture. These stress levels and their bases are defined in Paragraph 4.3.3 of the FSAR.
As a result of fast neutron irradiation in the region of the core, there will be an increase in the ND W with accumulated nuclear operation. The predicted maximum NDTT increase for the 40-year exposure is shown on Figure 4-10. (4)
The actual shift in ND W will be determined periodically during plant operation by testing of irradiated vessel material samples located in this reactor vessel. (5)
The results of the irradiated sample testing will be evaluated and compared to the design curve (Figure 4-11 of the FSAR) being used to predict the increase in transition temperature.
The desi n value for fast neutron (E > 1 MeV) exposure of the reactor vessel is 3.1 x 10 0 n/cm2 see at the reference design power of 2568 MWt and an integrated exposure of 3.0 x 1019 n/cm2 for 40 years operation.(6)
The calculated maximum 2
2 values are 2.2 x 1010 n/cm see and 2.2 x 1019 n/cm integrated exposure for 40 years operation at 80 percent load.(4)
Figure 3.1-1 is based on the design value which is considerably higher than the calculated value. The DTT value for Figure 3.1-1 is based on the projected NDM at the end of the first two effective full power years of operation. During these two years, the energy output has been conservatively estimated to be 1.7 x 106 thermal megawatt days, which is equivalent to 655 days at 2568 MWt core power. The projected fast neutron exposure to the reactor vessel for the two years is 1.7 x 1018 n/cm2 which is based on the 1.7 x 106 thermal megawatt days and the design value for fast neutron exposure.
The actual shift in NDTT will be established periodically during plant operation by testing vessel material samples which are irradiated by securing them periodically near the inside wall of the vessel in the core area to achieve an average effective exposure between 1 and 3 times that of the reactor vessel inner surface. To compensate for the increases in the ND W caused by irradiation, the limits on the pressure-temperature relationship are periodically changed to stay within the established stress limits during heatup and cooldown.
The NDTT shift and the magnitude of the thermal and pressure stresses are sensitive to integrated reactor power and not to instantaneous power level. Figures 3.1-1 and 3.1-2 are applicable to reactor core thermal ratings up to 2568 MWt.
Amendment No. 15
The pressure limit line on Figure 3.1-1 has been selected such that the reactor vessel stress resulting from internal pressure will not exceed 15 percent yield strength considering the following:
A 25 psi error in measured pressure a.
b.
System pressure is measured in either loop Maximum differential pressure between the point of system pressure c.
measurement and reactor vessel inlet for all operating pump combinations For adequate conservatism, in lieu of portions of the Operational Requirements of Appendix G to 10 CFR 50, a maximum pressure of 550 psig and a maximum heatup rate of 50 F in any one hour has been imposed below 275 F as shown on Figure 3.1-1.
The spray temperature difference restriction, based on a stress analysis of the spray line nozzle is imposed to maintain the thermal stresses at the pressuirzer spray line nozzle below the design limit.
Temperature requirements for the steam generator correspond with the measured NDTI for the shell.
REFERENCES (1)
FSAR, Section 4.1.2.4 (2)
ASME Boiler and Pressure Code,Section III, N-415 (3)
FSAR, Section 4.3.10.5 (4)
FSAR, Section 4.3.3 (5)
FSAR, Section 4.4.5 (6)
FSAR, Sections 4.1.2.8 and 4.3.3 1555 254 Amendment No. 15 3-5
4.2 REACTOR CCOLANT SYSTEM INSERVICE INSPECTION Applicability This technical specification applies to the inservice inspection of the reactor.
coolant system pressure boundary and portions of other safety oriented system pressure boundaries as shown on Figure 4.2-1.
Cbjective The objective of this inservice inspection program is to provide assurance of the continuing integrity of the reactor coolant system while at the same time minimizing radiation exposure to personnei in the performance of inservice inspections.
Specification 4.2.1 The inservice inspectien program to be followed is outlined in Table 4.2-1.
Except as provided for in this Table and as discussed herein, the inservice inspection program is in accordance with the ASME Code,Section XI, Rules for Inservice Inspectien of Nuclear Reactor Coolant Systems, dated January 1, 1970, as modified by the Winter 1970 Addenda.
Prior to initial plant operaticn a pre-operational inspection of the plant will be performed of at least the areas listed in the ASME Code, provided accessibility and the necessary inspection techniques are avail-able for each of these areas.
The only exception to this will be areas where the necessary base line data is already available t.nd has been obtained by the same techniques as will be used during inservice inspec-tion.
4.2.2 Reactor vessel irradiation capsules are planned to be withdrawn for testing at specimen exposures (E > 1MeV) eqsi'salent to 3, 9.5, 16, and 22.5 effective full power years of operation. Wi?.hdrawal schedules for testing may be modified to coincide with those refueling cutages most closely approaching the testing withdrawal schedule and may be adjusted following evaluation of data from each withdrawal in accordance with 10 CFR 50 Appendix H paragraph II.C.3.g.
Specimen capsules not subjected to destructive testing af ter Cycle 1 operation sdll be removed and stored during Cycle 2 operation, but shall be re-installed prior to Cycle 3 operation.
4.2.3 The accessible portions of one reactor coolant pump motor t?ywheel assembly will be ultrasonically inspected within 3-1/3 years, two within 6-2/3 years, and all four by the end of the 10 year irspection interval. However, the U.T. procedure is developmental and will be used only to the extent that it is shown to be meaningful. T4e extent of coverage will be limited to those areas of the flywheeA 9hich are accessible without motor disassembly, i.e., can be reached through the access ports. Also, if radiation levels at the lower access ports are prohibitive, only the upper access ports will be used.
1555 255 Amendment No. 15 4-11
300R OR M A k.2.h The inspecticn schedule =ay be =cdified to coincide with those refueling or =aintenance outages =cs: closely approaching the inspection schedule, k.2.5 Sufficient records 2r esch inspectier shall be kept to allow ec=parisen and evalur.icn of future inspections.
h.2.6 The inservice inspection shall be reviewed at the end of 5 years to consider incorpcratien cf new inspection techniques and equip =ent which have been preven practical, and a possible extension of the program to additional exa=inatica areas. The conclusions of this reviev shall be submitted to the AEC for evaluation.
Inspection 3r.ses The nuclear plant was designed prior to the issuance of Section a.
XI of the ASME Code, Rules for Inservice Inspection of Nuclear Reactor Ccolant Systens, dated January 1,1970.
Hevever, sur-ficient accessibility was included in the design to perfors = cst inspections discussed in the Code.
The preposed inspectica program follows the Code except that inspections are focused on areas vbich engineering analysis has indicated are subject to the
= ore critical stress, radiation, or transient ccnditions. The areas selected for inspection en this basis are listed in Table h.2-1.
These areas are exposed to the = ore severe conditions (which are still vell within Code limits) in the reactor coclant system. Therefore, they are expected to indicate potential proble=s bercre significant flavs develop in the selected areas or in other areas.
It is considered that the focused approach specified herein vill result in a =eaningful inspection progrs=
in that it vill provide assurance of continuing plant integrity.
In those areas where inspection methods are develop = ental, such as for re=ote inspection of the reactor vessel velds, reactor vessel nozzle inside radii and velds, and ultrasonic inspectice of pressurizer support bracket velds, the inspecticn methods vill be developed and tested to the extent practicable during pre-operational inspections. (Development of inspection techniques vill not be atte=pted en radioactive equip =ent, unless necessary to explore a specific probles.) A pre-operational inspection is planned of areas listed in the ASME Code which are withi= the inservice inspection boundaries and which are accessible for inspection. However, as discussed above, in areas where inspection
=ethods are develop = ental, the inspections will only be perfor=ed to the extent practicable. Once an inspection =ethod is selected for a particular inspection (e.g., U.T. for most volumetric in-spections), it is intended that all subsequent inservice inspecticns be performed using the identical =ethod and on the sa=e ec=ponent part' wherever practicable.
In additics to the above inspection, if any of the ec=penents within the inservice. inspection boundary are disassembled for =aintenance, the accessible parts will be given a nor=al visual examination as part of the routine plant =aintenance cperations.
b.
The vessel specimen surveillance program is based on specimen equivalent exposure years of 3, 9.5,16, and 22.5 EFPY referenced to 1/4 t(2).
These times were selected to meet the requirements of Appendix H to 10 CFR 50.
The specimen capsules not subjected to destructive testing after' cycle 1 operation are to be stored to permit the redesign of the capsule holders. The stored specimen capsules will be re-installed following completion of cycle 2 operation in a manner such that a specimen equivalent exposure (E > 1MeV) between 1 and 3 times that of the reactor vessel inner surface as required by 10 CFR 50 Appendix H is achieved.
The reactor coolant pump motor flywheel ultrasonic test procedure c.
is being developed to detect flaws of a small enough size to provide assurance of continued integrity, based upon a conservative fracture mechanics evaluation.
REFERENCE (1)
FSAR, Section 4.4 (2)
BAW-10100A February 1975 1555 257
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Amendment No. 15
f4 TA312 4.2-1 M
Og INSTRUMENT SURVEILIANCE REQUIREMDfTS She*L 1 of !!
O E3 item Examination Esamination p
IS - MI*
Category-IS - 2$18 Areas to be Emmained Methode Inspret ton Schedule and Futent itemarka
.vn
>3 Reactor Vessel and Closure Hemi GO gy 1.1 A
longitudinal and circum-Volumetric At or near the end of the 10 year hte 1**
ferential welds la core inspection laterval 105 of each gf3 reglon lorigitudinal and 55 of each cir-cumferential weld will be in.
]
spected. nosen the neutron flu-O ence esceeds 1019 avt (E 1 Mew E8 or greater) the length of each weld which is inspected will be increased to $O1.
1.2 B
longitudinal and circum-See remarks See Peserks The vende in this category SP ferential welds in the are the circumferential h
shell an3 hende (other weld just above the supgwrt than those covered in skirt junction with the s=
items 1.1. 1.3, and 1.3i) yeesel, the lower head to vessel welJ which in only acce.mnble from the L.ttua of the vessel, and the circumferential noas te twit weld.
11.e mtressen in theme wc1J:. are loves t h..
la ttw other reactor vene t welds to be insgeste4 per Items 1.1 as.4 1.1.
ft.cs c -
fore, no inspectiou= me e planneJ.
1.)
C Vessel to flange and teaJ Volumetric t/3 about every 3-1/1 years.
This in.nection will te to flange circumferential lerturned u, sang automateJ welds U.T. Nte I alp!!em.
C"3 g
13-261 and 10-251 refer to Tables in Section XI of the ASME Code.
w ** See the notes at the end of thia Table.
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UNITED STATES y*
4 NUCLEAR REGULATORY COMMISSION 3
1 WASHINGTON. D. C. 20665 k.....
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 15 TO FACILITY OPERATING LICENSE NO. DPR-50 METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER AND LIGKF CCMPANY PENNSYLVANIA ELECTRIC COMPANY THREE MILE ISLAND NUCLEAR STATION UNIT 1 DOCKET NO. 50-289 INTRODUCTION By letter dated March 23, 1976, Metropolitan Edison Company (the licensee) requested an exemption from the requirements of 10 CFR Part 50, Appendix H, Section II.C.2 to permit the operation of Three Mile Island Nuclear Station Unit 1 (THI-1), Cycle 2 with the reactor vessel surveillance capsules removed from the reactor vessel. The licensee requested corresponding changes to the Technical Specifications appended to Facility Operating License No. DPR-50 for the TMI-1.
These changes would reflect (a) the removal of the reactor vessel surveillance capsules for Cycle 2 operation (b) a revision in the surveillance capsule withdrawal times to conform with 10 CFR Part 50, Appendix H, Section II.C.3.C and (c) clarification of the term " effective full power years (EFPY)" to be consistent with standard Technical Specifications.
The licensee also advised that the capsule holder tubes would be removed for Cycle 2.
DISCUSSION The Three Mile Island Unit i design includes three reactor vessel surveillance capsule holder tubes located adjacent to the reactor inside vessel wall Each holder tube contains two surveillance capsules which hold the specaens to be irradiated in accordance with the requirements of the reactor vessel material surveillance program as described in Appendix H to 10 CFR Part 50.
The purpose of the surveillance program is to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel regions resulting from their exposure to neutron irradiation and the thermal environment.
1555 259
In a recent inspection of the surveillance capsule holder tubes, conducted during the current refueling cutage at THI-1, damage to the holder tubes was observed. He damage was attributed to contact and flow induced relative motion between the holder tubes and various corponents of the surveillance capsule train (push-rod spacers, holddown springs, and surveil-lance capsule rings) which position and hold the surveillance capsules in place during reactor operation. Two of the three holder tubes were found to be severed at the axial location of the second push-rod spacer from the top and one of these tubes was so 5everelyworn at the axial location of the first push-rod spacer that it became separated at that location during capsule removal. The third holder tube was intact following capsule removal.
To preclude the possibility of loose parts occurring during Cycle 2, the surveillance capsules and holder tubes will be removed prior to Cycle 2 operation. Engineering of new holder tube and push-rod assembly design modifications and material procurement will be completed during Cycle 2 to allow installation of the revised holder tubes prior to the start of cycle 3.
As a separate issue, the present schedule for withdrawals, subsequent to the first capsule withdrawal, is being modified to conform with Appendix H (which was issued after the TMI-1 Technical Specifications were developed).
In addition, Specification 3.1.2.4 is being changed to clarify that the first withdrawal would occur at a refueling outage nearest 2 EFPY of reactor exposure.
EVALUATION As required by Paragraph II.C.2 of Appendix H to 10 CFR Part 50, the surveil-lance capsules of TMI-1 are positioned during reactor operation so that the neutron flux received by the specimens is at least as, high, but not more than three times as high, as that recelyed by the vessel inner surface.
A recent calculation by the licensee, acceptable to the staff, indicates that the neutron flux is 2.4 times greater at the specimen location than at the reactor vessel wall at 1.4 vall thickness (1/4t). Cycle 1 of TMI-l accumulated approximately 1.3 EFPY of actual exposure to the reactor vessel wall at 1.4t; therefore the specimen accumulated approximately 3.2 EFPY of equivalent irradiation. Cycles 2 and 3 are planned to accumulate 0.8 EFPY and 0.75 EFPY of actual exposure respectively. Therefore, the specimens removed after Cycle 1 have already received an irradiation equivalent to 3.12 EFPY (1.3 x 2.4) which is more irradiation than the vessel wall at 1/4e will accumulate during the first three cycles of operation (2.85 EFPY).
1555 260
200R D H KL
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The irradiation effects accumulated during Cycle 1 will not be altered in those specimens stored throughout Cycle,2. When these specimens are reinstalled at the beginning of Cycle 3, the Technical Specifications will be revised, based on information acquired with the specimens tested after Cycle 1, to be applicable at least through the first 5 EFPY of reactor vessel exposure.
The specimen surveillance program is now based on equivalent exposure years of 3, 9.5,16, and 22.5 EFPY referenced to 1/4t so as to meet the requirements of Appendix H to 10 CFR Part 50.
In view of the above, we consider it acceptable to allow the licensee to remove the surveillance specimen capsules and holder tubes during Cycle 2 of TMI-1.
The specimen capsules not subjected to destructive testing after Cycle 1 operation may be stored until the beginning of Cycle 3 to permit redesign of the capsule holders. We also concur in the removal of the three capsule holder tubes during Cycle 2.
The withdrawal schedule that is based on specimen equivalent exposure years of 3, 9.5,16, and 22.5 EFPY referenced to 1/4t meets the requirements of Appendix H to 10 CFR Part 50.
He clarification of wording in Technical Specification 3.12 relating to effective full power years of operation is acceptable.
We have detemined that this amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that this amendment involves an.
action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR 351.5(d)(4) that an environmental statement, negative declaration, or environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
CONCLUSION We have concluded, based on the considerations discussed above, that:
(1) because the change does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the change does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
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1555 261
'[G[RGEL UNITED STATES NUCLEAR REGULATORY COMMISSION DOCKET NO. 50-289 METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER AND LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY NOTICE OF ISSUANCE OF AhENDFENT TO FACILITY OPERATING LICENSE Notice is hereby given that the U. S. Nuclear Regulatory Commission (the Commission) has issued Amendment No. 15 to Facility Operating Liceite No. DPR-50 issued to Metropolitan Edison Company, Jersey Central Power and Light Company and Pennsylvania Electric Company, which revised Technical Specifications for operation of the Three Mile Island Nuclear Station, Unit 1, located in Dauphin County, Pennsylvania. The amendment is effective as of its date of issuance.
The amendment provides for (1) the removal of surveillance capsules during Cycle 2, (2) the rescheduling of the surveillance program to conform with 10 CFR Part 50, Appendix H, and (3) the clarification of other requirements.
The application for the amendment complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the licease amendment. Prior public notice of this amende.ent was not required since the amendment does not involve a significant ha:ards consideration.
1555 262 eapmm
The Commission has determined that the issuance of this amendment will not result in any significant environmental impact and that pursuant to 10 CFR 351.5(d)(4) an environmental statement, negative declaration or environmental impact appraisal need not be prepared in connection with issuance of this amendment.
For further details with respect to this action, see (1) the application for amendment dated March 23, 1976, (2) Amendment No.15 to License No. DPR-50, and (3) the Commission's related Safety Evaluation. All of these items are availabic for public inspection at the Commission's Public Document Room, 1717 ll Street N. W., Washington, D. C. and at the Government Publications Section, State Library of Pennsylvania, Box 1601 (Education Building), llarrisburg, Pennsylvania.
A copy of items (2) and (3) may be obtained upon request addressed to the U. S. Nuclear Regulatory Commission, Washington, D. C. 20555, Attention:
Director, Division of Operating Reactors.
Dated at Bethesda, Maryland, this 14th day of May 1976.
FOR THE NUCLEAR REGULATORY C05SfISSION Vernon L. Roone, Ac ing Chief Operating Reactors Branch No. 4 Division of Operating Reactors 1555 263