ML20085C586
| ML20085C586 | |
| Person / Time | |
|---|---|
| Site: | Saxton File:GPU Nuclear icon.png |
| Issue date: | 08/02/1961 |
| From: | Neideg R SAXTON NUCLEAR EXPERIMENTAL CORP. |
| To: | |
| Shared Package | |
| ML20083L048 | List:
|
| References | |
| FOIA-91-17 NUDOCS 9110040024 | |
| Download: ML20085C586 (41) | |
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, SAXTON NUCLEAR EXPERIMENTAL CORPORATION Application for Renetor Construction Pemit and Operating Licence Docket No. 50-146 y i Dw oi C *r#3#
Amendment No. 9 In support of its pending application for an operating license for the Saxton reactor, Saxton Nuclear Experimental Corporation (SNEC) submits herewith, as attachments to this araendment, the following data relating to Saxton's financial qualifications.
Attachment A. Statements of actual and estimated project costs.
Attachment B. Balance sheet for SNEC as of June 30, 1961.
Atiachment C. Annual Report,1960 - Jersey Central Power and Light Company.
Attachment D. Annual Report, 1960 - New Jt..w y Power and 1* pt Company Attachment E. Annual Report, 1960 - Pennsylvania Electric Ccupany Attachment F., . Annual Report, 2960 - Metropolitas Edison Company Attachment G. Annual Report,1960 - Westinghouse Electric Corporation The foregoing material supplements and-brings up to date information contained in Section F cf Part A of the initial license application, including appendices to the license application referred to in said Section F, and infomation subsequently filed by Amendment No. 2 and Amendment No. 4 to the license application.
The material furnished with this amendment includes cost estimates for the Saxton project through December 31, 1962. This
-- period more than covers the estimated period required for completion 9110040024 910424 PDR FOIA Ftom CO. Hd %
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' Attachment _ L SAXTON llUOLEAR EXPERD* ENTAL CCRPORATION ACTUAL AMD ;TDIATED PROJECT COSTS Expenditures as of June 30, 1961 Paid to'destinghouse Electric Corporation in .
Accordance with Agreement Dated July 27, 1959 $ h,375,000 Application for Permits and Licenses 7h,8hD Training of Personnel 99,h50 Land and Site Preparation 15,87h Supervision and Engineering 68,5h9 Equipment, Supplies and Services 27,hh0 other Expenses 15,903
$ h,677,056 t
Estimate of Costs for Period from July 1, 1961 through December 31, 1962 Balance of Payments Due Westinghouse Electric Corporation in Accordance with Agreement $ 1,875,000 Permits anc' Liceroes 8h,000 Training of Perst .ael 98,000 hh, COO l Sup:rvision and Lngineering Operating Labor 200,000 Maintenance Labor 68,000 Spare Iarts and Supplies 120,000 i
Parts and Labor for Modification and Repairs of Existing Station Equipment 90,000 Insurance 115,000 Other Expense.o and Services h2,000 S 2,736,000 Energy Credit h6,000
$ 2,690,000
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2-of the reactor start-up program to be authorized by the initial prov1sional operating license, and allows sque roca for unexpected delays in that program. Cost estimates for periods subsequent to December 31, 1962, vill be fumished in connection with subsequent licensing proceedings.
The funds requind for the project through December 31, 1962, am amply covered by the amounts which bave thus far been budgeted and approved by the SIGC Board of Directors for the project by the four GPU domestic operating subsidiaries under the terms of their agreement with SNEC. They are also amply covered by the approvals which have already been obtained from State regulatory agencies and from the Securities and Exchange Ocamission for the transfer of funds by the GPU subsidiaries to SNEC.
SAXTON NUCLEAR EXPERDENTAL CORPORATION By /s/ R. E. Neidia (S E A L )
Attest:
/s/ E. L. Barth Secretary Svom and subscribed to before me this ' 2nd day of ' ,.
August, 1961.
(SEAL) /s/ Martin A. rnkr Notary Public
- ) Attachment A SAXTON NUCLEAR EXPERDI. ENTAL CORPORATION .
ACTUAL AND ESTIMATED PROJECT COSTS ,
Expenditures as of June 30, 1961 Paid to Westinghouse Electric Corporation in .
Accordance with Agreement Dated July 27, 1959 8 h,375,000 Th',8h0 Application for Permits and Licenses 99,h50 Training of Personnel L Land and Site Preparation 87h-15,5h9 Supervision and Engineering 68,hhD Equipment, Supplies and Services 27, Other Expenses 15,93
$ h,677,056 Estimate of Costs for Period from July 1, 1961 through December 31, 1962-l Balance of Payments Due Westinghouse Electric '
Corporation in Accordance with Agreement -
$ 1,875,000 Permits and Licenses 8h,000
[ 98 l Training of Personnel hh,000 Supervision and Engineering 3 000-l 200,000 1
Operating Labor Maintenance Labor 68,000
- Spare Parts--and Supplies 120,000. ,
Parts and Labor. for Modification and-
- Repairs of Existing Station Equipment 90,000' Insurance 115,000 l
. Other Expenses and Services- h2,000'
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8 2,736,000 h6,000:
Energy Credit-l
,$ 2,690,000_.
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SAXTON NUCLEAR EXPERDGNFAL CORPORATION BA1ANCE SHittT June 30,19(a iSSETS AND MHER DEBITS Construction Work in Progress 4,677,056 Current and Accraed Assets:
Cash and Working Funds 69.316 Total- Assets and Other Debits M LIABILITIES AND OTHER CREDITS-Common Stock (20,000 shares, $1.00 par) $ 20,000 Contributions fra Sharwholder Members ' L726.372 Total .
$4,746,372 Total Liabilities and Other Credits 84.746.372 b
Note: The Corporation has outstanding ocannitments and obligations -
under its contreet with Westinghouse Electric Corporation dated June 27,-1959 and as set forth in the proceedings-in S.E.C. File No. 70-3816 i
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DRAFT 25 Docket No. 50-146 CC*#'
9w oi Saxton Nuclear Experimenta1' Corporation Proposed License License NO.
- 1. This license applies to the pressurized water reactor.
(hereinafter referred tc as the " reactor")-owned by Saxton Nuclear Experimental Corporation'(hereinafter referred to as "Saxton"), located north of the Borough of Sa'xton in Liberty Township, Bedford County, Pennsylvania, and described in Amendment No. 5 dated April 19, 1961, and Amendment No. 7 dated June 30, 1961, to Saxton!s license application.
- 2. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses. Saxton:
A. Pursuant to Section 104.(b)-of,the Atomic Energy-Act ,
of 1954, as amended, (hereinafter referred to as :,he "Act") and Title 10 CFR, Chapter 1, Part 50, " Licensing of Production and Utilization Facilities", to' possess and use-the reactor as a_ utilization facility; ,
B. Purouant to the Act and Title 10 CFR, Chapter _1, ,
Part 70, Special Nuclear Material", to receive,-
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possess.and use _ kilograms of, contained Uranium.
235 as fuel for the operation of the reactor; and C. -Pursuant to the A.ct .and Title 10 CFR, Chapter 1,- Part 30, " Licensing of By-product Material" to possess,
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. but not to separate, such by-product material as may be produced by operation of thc reactor.
- 3. This .11censo shall be deemed i to contain and be subject to theconditionsspecifiedin5ection50.54ofPart50 and Section 70.32 of Part 70 and is subject to all applicable provisions of the Act and rules, regulations and ocders of the Commission now or hereafter in effect, and is subject to the additional conditions specified ,
below:
A. Technical Specifications The technical specifications contained in Appendix "A" to the license (hereafter referred to as the " technical specifications") are hereby incorporated in this license.
Except as .hereinaf ter provided, Saxton shall operate the facility only in accordance with the technical specifications. No changes shall be made in the technical specifications unless authorized by -the Commisation.
B. Authorization of. Changes and Experiments '
(1) Saxton may (a) _ make changes in the: facility as described in the: Final Safeguards' Report,-
(b) make fchanges in the procedures- described in
, the Final. Safeguards _ Report,:and (c)fconduct experiments-not described in the-Final Safeguards Report, unless. the proposed change or _ experiment
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involves a change in the_ technicai, specifications _ '
or an unreviewed safety question as defined in
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i subparagraph (2) below. Saxton shall promptly file with the Commission a report of each change or experiment carried out pursuant to the authorization granted in this subparagraph. If the proposed change or experiment involves a change in the technical specifications or an unreviewed safety question, it shall not be carried out unless authorized by the Commission pursuant to the procedures set forth in subparagraphs (3), (4) and (5) below.
(2) A proposed change or experiment shall be deemed to involve an unreviewed safety question if (a)
I the probability of occurrence of a type of accident analyzed in the Final Safeguards Report may be increased; or (b) if consecuences of any type t
I of accident analyzed in the Final Safeguards Report may be increased; or (c) if such change or experiment may create a credible probability of a nuclear accident of a different type than 1
any analyzed in the Final Safeguards Report.
(3) With respect to any change or experiment which must be authorized by the Commission pursuant to subparagraph (1) above Saxton shall submit a request for such authorization accompanied by an appropriate hazards analysis.
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- 4 (4) If the Director, Division of Licensing and Reg;1ation determines that the proposed change or (xperiment presents significant hazards considerations not described or implicit in ths Final Safeguards heport, he shall refer the request to the Advisory Committee on Reactor Safeguards and shcIl order a public hearing in accordance with applicable procedures. Saxton ,
shall be promptly notified of any referral to the Advisory Committee on Reactor Safeguards. If the-Director, Division of Licensing and Re5ulation, determines that the proposed change or experiment does not,prenent significant hazards considerations notdescribedorimplicitintpeFinalSafeguards Report, he may authorize such change or experiment without referr.al to the Advisory Committee on Reactor Cafeguards for a report and without scheduling a prior public hearing, upon finding that.there is reasonable aasurance that the health ; and safety of the public will not be endangered.
(5) Any report or request for authorization submitted-by Saxton and any1 determination or authorization-issued by the-Director,-Division of Licensing and Regulation, purauant - to subparagraph (4) above, shall be made a part of the public record of the licensing proceedings.
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5-C. Records In addition to those otherwise required under this license and app 31 cable regulations, Saxten shall keep the following records:
(1) Reactor operating records, including power levels and period of operations at each power level. ,
(2) Records showing the radioactivity released or discharged into the air or water beyond the effective control of Saxton as measured at or prior to the point of such release or discharge.
(3) Records of scrams, including reasons therefore.
(4) Records of principal maintenance operations involving' substitution or replacement of facility equipment or components and the reasons therefor, o (5) Records of radioactivity measurements atlon-site and off-site monitoring stations.
(6) Records of facility tests and measurements performed pursuant to the requirements of the technical specifications.
D. Reports, (1) In addition to. reports otherwise required under ,
this license and applicable regulations, Saxton shall make an immediate report in writing to tut Commission of any indication or occurrence of a possible unsafe condition relating to the operation
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of the facility, including, wichout implied limitation, any ace,1 dental rele.-se of radioactivity, whethe' Jr not resulting in personal injury or property damage or exposure above permissible limits, which affects the operation of the facility.
(2) Saxton shall immediately report to the Commission in writing any substantial variance disclosed by operation of the fJoility from performance specifications of the facility contained in the technical specifications.
(3) Saxton shall report to the Commissicu in writing significant changes in ope. rating procedures, plant organization, and transient or accident analyses, as described in the Final Safeguards Report.
E. Definitions (1) As used in this license the term " facility" means:
(a) The containment vessel which houses the reactor, eteam generator, main coolant system, other miscellaneous auxiliary systems, and the fuel ,
storage nell.
(b) The reactor core including the control rods, control rod drives, support structure, and normal operation instrumentation and controls.
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(c) The main coolant loop including the piping, steam generator, main coolant pump, reactor vessel, and normal operation instrumentation and controls.
(d) The pressure control and relief system which consists of a pressurizer, discharge tank, relief valve, safety valves, e3.etric heaters, and instrumentation and controls.
(e) The charging system conristing of high pressure pumps and instruments ap" rols.
(f) The purification system consitw 4 uent exchangers, flow control valve, demineralirers, and instrumentation and controls.
(g) The chemical addition system consisting of a steam heated boric acid tank, boric acid pump, and a chemical addition tank.
(b) The sampling and leak detection system consisting of high pressure and low pressure sampling and leak detection lines, aample coolers, sample a bombs, and instruments and controls.
(1) The shutdown cooling system consisting of a low pressure heat exchanger, pumps and instrumentation and controls.
(j) The safety injection system consisting of two high pressure pumps arranged in series, a high pressure and low pressure piping system, and instrumentation and controls, s
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(k) The station service electt'ical system eensisting of a normal and emergency power supply, 440-volt feeder busses, pressurizer heater control center, motor control centers, battery and battery charger, safety injection pumps supply, inverter bus and vital bus supply, and main coolant pump supplies.
(1) The radioactive waste disposal facility consisting of a solid waste disbosal system, a liquid waste disposal system, and a gaseous waste disposal system.
(m) The radiation monitoring system consisting of plant process monitoring, plant effluents monitoriag, site monitoring and plant area monitoring.
(n) Shielding incide and immediately outside the containment vessel, in the walls of and inside the containment vessel, in the walls of and inside of the control and auxiliary building, and in the waste treatt:e nt building.
(0) The fuel handling system consisting of special.
tools, hoists, and fuel storage rack.
(p) The secondary coolant system inside the containment vessel and the piping outside the containment vessel up to the feed water regulating valve and the steam pressure regulating vclve.
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(2)- As used in-this license, the term " Final Safe-guards Report" means the report so' designated by Saxton and aubmitted by Amendment No. 5 to ,
Sexton's license application, dated April 19,.
1961; and includes the supplemental information submitted by Saxton in Amendment No. 7, dated June 30, 1961.
- 4. A. Operation of the facility shall not begin until the-Director, Division of Licensing and Regulation, has found that construction of the facility .. bas.been completed in conformity with the Final Safeguards Report; I provided that the first core.may be loaded and the facility operated with the reactor ' head off at a power level not exceeding 100 thermal kilowatts prior to a finding of completion of the following systems:
(11 Pressure control and relief system.
. (2) Purification system.
(3)' Sampling and leak: detection-system.
(4) Safety-injection system..
(5)_ Radioactive l waste disposal system.
(6)- secondary coolant system._ ,
(7) Radiation' monitoring for_ the steam generator shell blowdown and the liquid effluent from the waste-treatment plant.-
(8) -Power supply for the above-systems.
(9) outer door of normal and emergency personnel access openings to. containment vessel.
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B. A copy of all findings pursuant to Paragraph A shall be promptly placed in the. record of the licensing proceedings.
5 .. This license is effective as of the date of issuance and shall expire December 31, 1962 (unless extended for good cause shown), or upon the earlier issuance of a superseding ,
operating license pursuant to further order of the Commission.
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APPENDIX A SAXTON NUCLEAR EXPERIMENTAL CORPORATION TECHNICAL SPECIFICATIONS
- 1. The facility is located ca the Saxton Steam Generating Station property of the Pennsylvania Electric Company near the Borough of Saxton, Pennsylvania, in Liberty Township, Bedford County, Pennsylvania. The Pennsylvania Electric Campany property consista of approximately 150 acres along the Raystown Branch of the Juniata River and the minimum distance from the center of the containment vessel to the nearest boundary including the river is 800 feet. The principal activities carried on within this property are the generation and transmission of electric power by the Pennsylvania Electric Company and the preparation and sale of ashes from an ash dump located at the eastern end of the prope.ty.
- 2. The steady state reactor power level vill not exceed 20 MWT.
3 The containment vessel design pressure is 30 peig and the maximum total leakage rate including penetrations is 0.2% of the free volume 2 in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at design pressure. A test to confirm the containment vessel and penetrations leakage rate shal.1 be made at not less than 10 psig within cme year from the time the fuel tenn operating license is issued. The frequency of future tests vill depend upon the results of this initial test'and consultation with the AEC Hazards Evaluation Branch.
The personnel entrances to the containment vessel have double door air locks that are mechanically interlocked to prevent both doors from being opened at the same time. A mechanism requiring a special tool or key
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-2 is provided to override this interlock under certain ccaditions which t are given under maintenance operating procedures in these technical l
specifications.
The containment vessel dimensions, materials of construction, frge .
vol me and containment vessel penetration information are given in Subsection 223 of the Final Safeguards Report (Figures 201-3, 201 h, 201-5, and 201-6 and subsection 506 that an referred to in Bubsectica 223 are not included as a part of this technical specification.)
Provision has been made for a total of 325 electrical penetrations and 87 piping penetrations including spares. There are approximately 150 spare electrical penetrations and 40 spare piping penetrations.
During operation when the contairanent vessel is closed, the internal pmssure vill be maintained at essentially atmospheric pressure. A containment vessel pressure alarm set for 5 psig vill be provided in the main control room. Valven in the main steam line, Ventilating purge lines and purification system lines can be closed remote manually l from the main control roca. Fai.1 safe valves in the lines between the fuel storage well in the contairnment vessel and the refueling water storage tank vill be automatically closed in the event the pressure in the containment vessel exceeds 5 psig. The valves in the ventilation purge lines vill be in the closed position when t5e reactor is operating at power.
- 4. One main coolant system consistdag of one Type 304 stainless steel piping loop, one reactor vessel, one canned rotor pump, one steam generator, and one p2tssurizer are being provided. The reactor vessel design features
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types of connections and number of penetrations are given in paragraph B on page 204.2 of the Final Safeguards Report.
The main coolant is demineralized light water containing boric acid an required for shutdown reactivity control. The maximum permissible activity of the main coolant shall not be more than 20 ue/cc of lon6-lived isotopes. The main coolant shall make ore pass throup,h the core in an upward direction. The main coolant operating limitations.under steady state conditions are as follows:
(a) Minimum inlet pmssure 1900 psia (b) Minimum flow rate 6200 gp (c) Maximum mixed core exit temperature 5600 F An audible and visible ala m system in the main control roca vill alert the operator when the above operating limitations are reached.
The principal featun s of the major components are as follows:
(a) The steam generator is a vertical shell and U-tube type vith integral--
steam drum and thne stages of moisture separation. This steam generator is designed with 40% excess capacity above the nominal 20 WT rating of the core.
(b) The =av4== primary safety valve setting is 25T5 psia.
(c) The total rated capacity of the pressun relief system'is 65,000 lb/hr of-saturated steam.
(d) The purification system is rated at 30 gp and will maintain the impurity level of the main coolant water at-less than 1 pp with a flow rate of-10 gp. The main coolant impurity level vill not-be- allowed to exceed 5 ppm solids.
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-u-(e) A sampling system is being provided to periodically sample main coolant on the discharge side of the primary coolant pump, pressurizer water, and inlet and outle ; of the purification and boric acid demineralizers.
(f) The primary shield is described in Subsection 221 of the Final Safeguards Beport. (Details shown on drsvings 201-3, 201-4, 201-5, 201-6, 201-7, and 201-8 referred to in subsection 221, except for shield dimensions, are not a part of this technical specification.)
5 The secondary coolant is light water and steam. The operating pressure vill be approximately 605 psia when transferring 20 MWI. The maximum permissible pressure upstream of the pressure reducing valve is 1800 psia. The == vi == temperature vill be the sature, tion tempers.ture for the steam pressure developed. The steam flow rate vill be approximately 69,000 lb/hr at 20 Mwr. -
- 6. The reactor core having the following features is provided:
(a) The main coolant, which is light water, vill serve as the moderator and reflector. The effective reflector thickness is 10 inches.
(b) Uranium oxide (UO 2 ) enriched to 5.% U-235 vill be used for fuel.
Uranium oxide (UO 2 ) at the density being used has a melting point of epproximately 5000 0 F.
(c) Type 304 stainless steel ir used for the cladding of the fuel assembly rods that ztquire no further veldin6 or brazing. Type 3k8 modified carbon stainless steel is used for the L shaped fuel assemblies. The melting point of stainless steel is epproximately 2550 F.
(d) The fuel element description and dimensions are given on pages 203 1, 203 2 and 203 3 of the Final Safeguards Report. . (The' number of fuel i
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5 rods in the core, the numM r of fuel pellets per fuel rod, fuel pellet and pellet column tLbeension tolerances, fuel rod end gap dimensions, and the total weight of UO 2 in the reactor core as given on these pages are not a prt of this technical specification.)
(e) The =mv4=m total weight of fuel in the core is approximately 2250 lbs of UO2 enriched to 5 7 weight per cent U-235 isotope.
(f) The maximum number of fuel assemblies in the initial core is 21.
(g) The maximun fuel burnup will not exceed 30,000 MWD /MTU.
(h) The calculated void coefficient of reactivity at operating temperature villnotexceed-0.14%k/f, void.
(1) The temperature reactivity defect is approximately 8.2% k from cold clean to hot clean conditions at zero power.
(j) Boric acid vill be injected into the main coolant system to maintain the reactor suberitical ty at least 3% k when all the control rods are inserted and the main coolant temperature is less than 4300 F.
During initial core -loadig, refueling, _ or during any movement of the control rods or fuel assemblies with the nactor vessel head off, sufficient boric acid vill be injected to maintain the com suberitical by 10% k with all contrM rods inserted.
The boric acid system cor siscs of a 50 cubic foot capacity steam-heated tank for_p nparati.cn of the concentrated boric acid and a 30 gpm stainless steel boric acid pump and piping for tmnsferring the boric acid to the charging system which is used to inject the boric acid into the main coolant-system.
The boron concentration :1n the main coolant when the reactor is critical at zero power lovel vill not exceed 200 pga. The reactor
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vill not be brought to full power until the boron concentration haa been reduced to less than 10 ppm.
(k) The principal core temperatures and thermal characteristics at 20 MVI under steady state conditions are ab follows:
Nav4"n'= calculated heat flux, Btu /hr-ft2 444,000~
Average heat flux, Btu /hr-ft2 137,000 Minimum burnout safety factor (ModifiedBettisCorrelation) 2.4 O
Maximum calculated fuel clad temperature. F 642 Maximumfuelheatgeneration,XW/ft 13 3 Average power density, KV/ liter, core 54 (1) Under credible accident conditions as described in the Saxton Final Safeguards Report, the minimum calculated bumout safety factor is not expected to be less than 1.85 7 The reactor control and safety system haa t 6e following features:
(a) There vill be a total of 6 offset cruc6 form-shaped control rods. '
t The minimum number of operative contrr. rods is six. The control rods and control rod drive mechanisms a m described in paragraph D of Subsection 203 of the Final Safeguards Report. (Details'shown on Figures 203-5a and 203-5b referred to in this paragraph are not part of this technical specification.)
(b) The approximate reactivity vorth of the six control rods at an operating temperature of 530 0 Fis24%k.
(c) The =mvinnmi reactivity vorth of any control rod group vill be approximately12%k.
.(d) The minimum scram control margin at operating pressure and temperature with one control rod stuck vill always be at least 2% k.
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-7 (e) The maximum reactivity addition rate if all six control rods could be withdrawn at operating temperature at the =v4== vithdraval rate
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of 15 inches per minute is approximately 5 x 10 k/sec. An interlock (see item h) is provided to limit the reactivity insertion rate. The
= vi = = reactivity addition rate,if the boric acid concentration is reduced by means of the bleed and feed system operating at the maximum .ste of 30 gIrn,is approximately 4.7 x 10-5 k/sec.
(f) The maximum excess reactivity when the reactor is cold clean is ,
approximately 25% k.
(g) The reactor vill be automatically scrammed under the following enaditions:
Conditions Bet Point Fast startup rate (Navi= =) -
2 decades / min High power level at startup (Mari==) 5 MW HiGh power level at power (Maximum) 24 MW Lov main coolant pressure (Mini ==) 1600 psig Lov main c.colant flow (Minimum) 2.2 x 106 lb/hr Lov vater level in pressurizer (Minimum) 10 inches Loss of main coolant pump power Contact en breakers High main coolant temperature (Hot Leg) (Maxinnu) 6000 F (h) The following interloe.ks are provided:
g Function Condition of Use Electrical Limit the =vi== reactivity Automatic during all (Adjustable resistors) insertion rate to 2 5 x 10-4 'sperations Electrical Limit rod withdrawal to Adjusted periodically (Relay on position in- provide 2% shutdown margin based on rod worth dict. tor coils) with one stuck rod, curves snd operating conditior.s.
Electrical Remove startu;. rate rod stop Automatic during and scram above 10% power, startup i
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M Function Condition of Use 1 Electrical Preveat automatic operation Automatic during startup of control rods belov 10% and drops in power power 1
- 41) The following by-passes or overrides are provided:
By-pass Method Condition of Use Source range Manual switch in main control Power operation channegsabove room 5 x 10 nv Power range Manual switch in main control Belov 10% power and channel coinci- room when any power range dence channel is not operating Startup power Manual switch in main control When power level level scram room exceeds 10%
Lov main coolant Push button in main controi During startup or or pressurizer room (t. elf reset above 1600 cooldown pressure psia)
Safety injection Manual switch in main contro] During startup, cool-room down or testing of safety injection pumps (j) The range of use of the startup (intermediate range) 2ste scram 0
-is 2 5 x 102ny to 7 x 10 nv where 7 x 10 0ny is equal to 10% of full power (20 NW).
(k) The maxima total scram time from scram initiation to scram completion vill not exceed 1 5 seconds.
(1) The minimm number of operative power level safety channels is 2.
The range of use of power level safety channels is from 1% to 150%
of full (20 MW) power. These channela are in service during startup as well as during power operation. The mini == number of startup rate channels is 2.
(m) Three power level channels am provided and coincidence must exist between at least two channels in order to scram the reactor. A
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switch is provided for overriding this coirch 3:e when below i 10% power and when any power range channel is out of service.
(n) An automatic and manual control system is Iror:Lded for positioning the control rods and regulating reactor output.. 7.,ach of the control rods has an automatic-off-manua?. control suitch arid the rods can be operated individually or in groups. The aittomatic control system is designed to handle changes of approximately 2 F.W per minute; however, the pressure control sys1em is deilignod f.o handle full load rejection without opening the safety valvos.
B. A standby power source to the reactor station w.11 he provided through a 750 KVA, 2300/hh0 y transformer from the 2300 y station service busses in the existing plant. In case of power failure from the normal power supply, which is a 1000 KVA,13.2 KV/hhD v tes.nsformer, an automatic transfer will close a breaker picking up the standby power source. A 125 v d-c battery and an inverter-diverter is provided to supply power to an a-c bus used for the vita.1, ir.struments which include nuclear instrumentation, radiation monitoring, pagir:g system, rod control and position indication and all process a-c instrurentation and control from the main control room.
- 9. A radiation monitoring system having the following significant features is being provided:
(a) The air discharged into the plant stack is cercsinuously monitored by beta-gamma detectors mounted in a selded stsel pipo frants l inside the duct leading to the exhaust fan in the bas O the stack. The threshold sensitivity of these detecters is 2.0 x 10'I-1 l uc/ce. Any activity detected by this monitor is recorded and alarmed in the main control room.
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(b) The iodine activity in the main coolant is monitored by a gama detector located in the bleed line after the non-regenerative heat exchanger. The sensitivity range of this detector is 2 x 10'N to 2 x 10'l uc/ccforenergiesabove0.8mev. An alarm is provided in the main control room to alert the operator, if this activity exceeds the background activity by 25%.
(c) Detectots are provided to continuously monitor the containment "
vessel air for radioactive particulate matter and radioactive gases. A small amount of the containment vessel air is continu-ously drawn through a section of moving filter paper by a constant displacenent pump. A detector that has a sensitivity range of 10~9 to 10-6 ue/ccmonitorsthefilterpaperaorbetaradioactivity from particulate matter. A detector that has a sensitivity range of 3 x 10-6 to 3 x 10-3_ue/ccinabackgroundof0.6mr/hr_ con-s tinuously monitors the air withdrawn from the containment vessel.
Alarus are provided in the main control rom to alert the-operator, if-the particulate monitor radioactivity or the gas monitor radioactivity exceeds 25% above background.
(d) Six plant area-shelf-moun w monitors are provided. These monitors am located in the vaste treatment plant control rom, the reactor plant main control ro m , the charging room, the sampling room,- the chemical preparation -1sberatory,. and the health Ihysics' office. All of these monitors have a sensitivity range of 0.01 mr/hr to 10 ar/hr, except the monitor in the main control room wMeh has-a sensitivity of 0.2 ar/hr to 200 mr/hr. 4 The plant shiel q has-been designed to limit the-radiation-levels in-the various areas approximately as follows: '
T/26/61 Continuously occupied areas (such as_ main control room) 0.25mr/hr s.
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Intermittently occupied work areas (such as var +
treatment building control room) 075mr/hr Intemit t%tly occupied ground level arca-duringoperating(suchasareasadjacentto containmentvessel) 2.0 mr/hr Containment vessel during refueling operations 2.0 mr/hr Additional shielding vin be added if mquired to nouce radiation levels M the above areas below those specified in AEC Begulation ,
10 CFR ru ;, 20.
(e) A ganna detector that has a sensitivity of 0.01 to 10 mr/hr using Cobalt 60 as a referen*e is provihd to continuously monitor the stnam generstor shell side blowdown water to detect any leakage frva the main coolant system. An alarm de provided in the main control room to alert the operator if the, rad.ioactivity of the blowdown exueeds 2 x 100 uc/cc. A proportiinal type sampler also continaously samples the discharge fmm the steam generator blevdown tank and a sample vill periodically be taken to the laboratory for counting.
- 10. A radioantive vaste disposal system having the fonoving featuns is being provided:-
(a) A purification system having an s.d;ostable flow rate from 3 gpn to 30 gpn vin continuously remove radioactive particulate matter frcss the nain coolant. This radioactive particulate matter vin be concentrated on ion exchange Mains which vill periodicany be transferred to underground storage tanks. Radioactive gas nieased to the purification system surge tank if11) be removed by tvc rotary ater-sealed type gas ccanpressors and v3.11 be temporarily stored in three 133 cubic feet gas decay tanks before being released at a con-7/26/61 tron ed rate to tha stack.
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(b) Radioactive particulate matter vill be continuously removed from the water in the reactor and stomge vell compartment by means of ion exchange resins. These resins vill also he periodically trans-ferred to underground storage tanks. The nominal flow rate for this system is 15 gpn.
(c) The stack for discharging air-borne radioactivity is 125 feet in height. The exhaust fan discharging into this etw k has nominal capacity of 25,000 cfm.
(d) Two 10,000-gallon tanks and one 5,000-gallon tank an provided to store radioactive water discharged from the main coolant system or radioactive contaminated water from other various sources within the plant. These storage tanks are horizontal, cylindrical tanks mounted inside a second horizontal, cylindrical tank. The tank within a tank debign provides an ennular space for detecting and monitoring leakage. Three 800-gallon tanks of the dual type construction vill be provided for storing the spent demineralizer resins, P.ree 133 cuts feet capacity tanks are provided for the g rasloactive gas collected by the gas compressors. All of these tanks are buried underground.
(e) Radioactive liquids vill be processed through a vaste treatment plant consisting of a gas stripper and an evaporator unit, each naving a nominal feed capacity of 1,000 pounds per hour. All of this equ11 :nent is constructed of Type 304 stainless steel and the evaporator includes an evaporator still pot, demister, and vent cendenser. Concentrates from the evaporator vill be combined with cement inside various size standard steel drums. A vood crib and C
carth shielded storage area is provided for storing the filled 7/27/61 dnans prior to shitment by an AEC licensed carrier.
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(f) No radienetive colid vaste vill be dit. posed af at tht site.
(g) No radioactive liquid vaste concentrations in excess of those specified in Appendis "B", Table 71,10 CFR 20, vn1 be discharged to the river.
(h) Corrective action vill be taken to keep the activity concentration of gaseous radioactive releaus, as measured by the radiation monitor in the ventilhting duct ahead of the stack discharge fan, less than 2 x 10*7 ue/cc. By controning' the stack concentration so that it does not exceed 2 x 10-7 ue/ce, the maximum ground level concentrstion averaged over a period not_ exceeding one year vill not be in excess of the limits specified in Appendix "B",
Table II, Column 1 of 10 CFR 20.
4
- 11. Ihnergency cocling systems are provided for the following conditions:
(a) In case of total power nupply failure to the nuclear plant, emergency cooling vill be provided for by natural circulation. -
Steam-driven boiler feed pua.ps make it possible to intemit-tently supply water to the steam genemtor for removing heat frosn the main coolant system.
(b) In case of loss of the steam geerstor as a heat sink, emergency cooling can be provided by means'of the parification system and charging system. A by-pass valve, tha.t is remotely controlled i
from the main control rom , is located in the high pmssure main.
coolant line to the regenerative heat exchanger. This valve makes it possible to by-pass the regenerative heat exchanger and amove-heat froan 'he main coolant system by means of the non-n generative heat exchanger and- the caponent cooling system. During this
-7/27/61 period, the flow rate can be increased to at least 30 spa and a
)
total of approximately 1,800,000 Btu per hour can be removed.
(c) In the event of a rupture of any part of the main coolant system and a resultant loss of coolant, a safety injection cystem is provided to keep the core covered vith water containing approxi-mately1%boricacid. This system consists of two 375 gym motor-driven centrifugal pumps piped to operate individually or in series and discharging into two separa?,e 3-inch pipmlines feeding directly into the reactor vessel. Borated water for these pumps is supplied from an 80,000-gallon heated storage tank located in the yard adjacent to the containment vessel. These pumps ce.n be started manually from the main control rman or vould be started automati-cally in the event the main coolant system pressure falls to 1,000 pai or less during operation. A flow control channel for each of the two lines feeding the reactor vessel is mounted on the main control hoard. In case one of these lines ruptures, the flow in the line vould become abnormally high. This v.*ll be detected by the flow control channel and an associated control valve vill be autcunatically closed, thus diverting e flow to the remaining line.
- 12. A charging system consisting of two motor-driven horizontal, triplex, reciprocating pumps, each having an adjustable capacity of 3 to 15 spm at a total dynamic head of 2500 psig, is provided to charge degasified, demineralized make-up vater, concentrated borated water, and water from the purification system back into the main coolant system. One of these pumps vill operate continuously for the purification system and for volume control of the main coolant syntesa.
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13 A shutdovn cooling system is provided for removing decay heat after the main coolant system pressure has been reduced to 150 psig or less. This system consists of a heat exchanger and two 50 gpa cen-trifugal pumps. Only one pump is required to remove the decay heat and the second punp is provided for use as a spare in the event the other pump is shut down for maintenance. This system also has a spare heat exchanger in that the purifiention system non-regenerative .
heat exchanger can be used as a spare in the event the shutdown cooling system heat exchanger is out of service fo** tepairs.
- 14. Provisions have been made to test the major safety systems and controls.
These systems and controls vill be teated in accordance with the following schedule:
Safety Injection Ptraps and Automatic Startup Control Monthly Radiation Monitor Circuits Monthly Control Rod Drive Scram Speed Every 3 mmthe Serra Settings Every 6 months Scram Circuit Response Time Measurement Every 6 months 15 The No. 2 turbine has a manual trip and an overspeed trip. The manual trip is normally used every time the tarbine is shut down and the overspeed trip vill be checked st least every six months.
- 16. The following categories of core parameter measurement tests vill be made as . coon as possible after initial core loading and prior to increasing reactor power above 2 MWP.
(a) Control rod drive and eczum test (b) Criticality with banked and progresser.d rods 7/27/61 (c) Control rod, boron, flow,. and simulated void worth determination
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at ambient tenperature (d) Temperature and pressure coefficient at shutdown boron concentration (e) Control rod vorth, boron, and flov vorth at operating temperature (f) Temperature and pressure ccefficient deterrtination at low boron con-centration (g) Instrumentation response at lov power 17 The following categories of power operational tests vill be made following completion of the tests outlined in paragraph 16 above.
(a) Power coeffici; J.t measurements (b) Loss of load transient tests (c) Calibration of Nuclear Instrumentation System and Primary-Secondary Calorimetry (d) FissionproducteffectonreactivityfollowingpowerincreaseasS/or decrease (e) Biological shielding effectiveness test (f) Instrumentation and control response (g) Cooling by natural circulation
- 18. Administrative and procedural safeguards:
(a) Detailed vritten procedures vill be prepared and vill be put into effect for all nomal operations, maintenance operations, and emergency operations which any affect nuclear safety. ,.
(b) Administrative provisions and controls for maintaining nuclear t
safety are:
(1) Records suen as log books and log sheets vill be maintained for all normal operations and tecting ope mtions. System check-off lists and instrument and control set point lists 7/27/61 vill be maintained in the main c % trol room at all times.
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All important operating records, including recorder charts, records of radioactivity releases, and any other records required by permhi or licenaes, vill be kept on file for L
i the life of the prcject.
l l (2) An AEC litansed operator vill be in the main control room I
i r.t all times while there is fuel in the reactor, with the
( exception of periods when no reactivity changes are being made and the reactor is at least 5% suberitical at ambient temperature with the scram circuit breaker locked open.
(3) In the event of a reactor scram, the reactor 5.11 not be t
3 started up again until the action that initiated the scram can be ascertained and corrected. If the trouble cannot be located imediately or if corrective action cannot be taken by means of supervisory instntments and controls in the main control rocxn or equipment that is accessible outside the containment vessel, the Supervisor of Operations and Tests or the Nuclear Plant Superintendent vill be notified imediately.
One of these supervisors, or a supervisor of equivalent qurtli-fications, vill always be on call for such emergencies.
(4) Fueling, refueling, and the installation or removal of experi-mental fuel assemblies or control rods vill be supervised by qualified Westinghouse and/or Baxton Nuclear Experimental Cor-poration personnel.
(5) All operations, that could in any way affect the safety of the plant, vill have to be cleared with the Reactor Plant Supervisor on duty.
(6) Operation of the Unit No. 2 turbine-generator and its auxiliary 7/27/61 equipment, as well as other nuclear plant facilities located in
_ _ _ _ _ _ _ _ - - _. a
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the existing Pennsylvania Electric C mpany power plant, vill be under the jurisdiction of the Sexton Nuclent Ex-perimental Corporation.
(7) Before any major circuit breaker or critical valve can be operated or any najor piece of equipnent, control, or in-strementation can be taken out of service, a written order must be issued and signed by or on authority of t're Reactor Plant Supervinor on duty.
(8) A system of radiation work permits will be used to control access to areas where radiation overexposure of personnel is possible within a nomal work day. Radiation monitoring and assistance by radiation protection yersonnel vill be provided as required to minimite exposums.
(9) Radiotetiva liquid effluents and radioactive gaseous effluents vill only be discharged to the natural environment upon instructions from the Supervicor - Reactor Plant Services or his designee. The discharge of these effluents vill then be supervised by the Reactor Plant dupervisor on duty.
19 Principal operating procedures having a potential effect on safety are listed below:
(a) Initial core loading:
(1) An AEC licensed operator vill be present in the control rom.
Personnel adequately trained in the fueling operation vill also be present in the containment vessel to supervise this operation.
(2) Sufficient boron vill be added to the main coolant to render the core 10% suberitical with a fully loaded, rodded core.
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f (3) One control rod g,roup vill be cocked for nafety purposes. l l
(4) Once the addition of fuel to the ccre has begun, no opern-l.
tion vill be performed which will reduce the boron concen-
! tration in the Tin coolant systen.
(5) If at any time the experimentally extrapolated value for the critical size of the reactor core (as indicated by the 1/M datafromanydetector)islessthantwofuelassemblies, plus the number of fuel assemblies then in the core, suffi-cient boron vill be added to satisfy this condition.
(6) Four BF3 channels will be used for loading; of these, three '
i must always be functional.
(7) No more than one nuclear detector vill be relocated between '
v any two dats steps.
(8) Should the loading operation be interrupted for a period of l
l more than two hours, a new set of data vill be taken before-a nev loading adjustment is.ande. If the new data deviate by more than 20% from the corresponding previous data,- the last loading step involved in the addition of fuel vill be repeated and the extrapolation to criticality checked. .
(b) Initial criticality:
i j (1) Nuo' instrumentation required vill have been installed j .-
and aceked out.
(2) Centrol rod configuration for criticality will bave been estimated.
(3) Startup ru1,es vill not exceed one decade per minute.
(c) Initial core parsmeter measurements: #
(1) Any plant changes which would produce a sudden lowering of;
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reactor coolant temperature (of the order of 10 F) vill be avoided while t).e reactor is approaching criticality or is critical.
(2) Available shutdovn reactivity vill not be reduced belov 1% k vith a stuck control rod as detemined by previous analysis and experiment. That is, the bomn concentration in the system vill not be reduced below the value required to give the keft of approximately .99 for the core if one control rod should be stuck in a preset safety position vith the others inse rted .
(d) Reactor startup:
(1) All personnel vill be notified whenever startup of the reactor plant is iminent.
(2) All control rod position indicators vill be checked to ascer-tatu that the rods are in the fully inserted position.
(3) The follering minimu:n startup instrument requirements vill be cet:
(i) Two source channels (ii) Two intermediate range channela (iii) Two power range channels (k) Pressurizer heatup rate vill not be ellowed to exceed 250 F per hour.
(5) Differential between main coolant temperature and pressurizer temperature vill be maintained less than 2000 F and pressurizer sprays vill not be used if the differential exceeds 2000 F.
(6) The main coolant pressure vill not be allowed to exceed 100 psig until the temperature of the main coolant is at least 2600 F.
7/27/61
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. N (7) When bringing the main coolant system up to temperature and pressure, boron removal vill not be started until the main coolant syster temperature reaches 2500 P. ,
(8) The safoty injection controller vill be evitched to auta::Atic position when main coolant system pressure exceeds safety inje: tion set point pressure by 200 psig.
(9) When main coolant system reaches operating temperature of 5000 P, the approach to criticality vill 1< ~'^ eted by inter-mittent program rod withdraval. The rod progma vill be adjusted so as to assure that there is at leaat 24 shutdown upon scram with the most effective rod stuck.
(10) During approach to criticality, flux multiplication rates vill be maintained at less than 1 decade per minute.
e (e) Beactor operation:
(1) In a condition where one power range channel is inoperative, the reactor can continue to be operated at the existing power level provided no uajor rod program change is carried out and provided the power range high level scram protection is set for single channel operation.
(2) Periodic calorimetric determination for both primary and seccndary systens vill be performed and the nuclear power channels vill be adjusted if necessary.
(3) All annunciator alarms vill be acknowledged prcasptly and abnomal conditions will be corrected as soon as possible.
(4) If the normal or emergency power supply is lost, the reactor will be shut down antil such time as both power supplies are in service.
. 7/27/61
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(5) Waste dispocal storage tanks vill be arranged to receive water ejec+ed free the main coolant nystem.
(6) The neutron flua level instruments vill be monitored as the station load is changed.
(7) When the reactor is critical at tero power level or when the load is being mduced to zero power level, secondary s, team pru aure vill b8 adjusted to he.intain the main coolant system above 4300 F.
(f) Reactor hot shutocr.nt (1) One control rod group vill be left in at cocked position ,
equivalent to a reactivity of approximately 1% k.
(2) The main coolant system vill be usintained at operating <
temperature and pressure by means of main coo.1. ant pump heat, maidual heat in core, pressuriter heaters, and dischar6e of stews to main steam header.
(3) A minimum of one eenree range flux channel and one inter-mediate flux enannel vill be in service.
(g) Reactor cold shutdowns (1) One control rod group vill be left in a cocked position equivclent to a reactivity of approximately 1% k prior to reducing the temperature and preocure.
(2) The cooldown rate vill not exceed 2000 F per hour. g (3) When the main coolant temperature reaches 4300 F, a sufficient amount of concentrated boric acid vill be injected to make the reactor suberitical by at least 5% k at ambient temperature with all rods inserted in the com . After i suitable mixing 7/27/61
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time has elapsed, the boron content of the main coolant vill be determined by sempling.
(4) The "lov main coolant pmsaure" scram override vill be initiated prior to reducing the main coolant prersure.
(9) The safety injection control switch vill be placed on manual when the pressure reaches approximately 1000 psig.
(6) The shutdovn cooling system will be put in operation afte.+
the main coolant system reached 300 0 F and 150 psig. This system vill M kept in operation or operated intemittently as needed to keep the main coolant temperature at or below 140 F.
(7) A miniava of me source range flux channel and one inter-mediate flux channel vill be in service.
(b) Maintenance:
(1) Only authorir.ed personnel vill be allows a to enter the con-i tainment vessel to perfom insp1ctions end maintenance work.
When the nactor is in a hot shutdown conditita, at least two men-vill be sent into the containment vessel whe: sever it -is necessary to make inspections or perfom minor repairs.
M:nor repairs vill censist of instrument and equipnt ad-justments that can be made with amm11 h ud tools.
(2) A minimum of one source range flux channel and one inter-mediate flux-channel vill be in servico.
(3) Radiatinn monitoring equipnt and ~alaras vill be in service.
, (4) The various areas of the plant and comprtments in thc. con-tainment vessel vill be monitored for radiation and contamin--
7/aT/61
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- 2'4-ation prior to carrying out maintenance work.
(5) L as and syskens vill be sampled for radioactive gases prior to being opened up for maintenance.
(6) Special clothing vill be provided for maintenance work to prevent or n strict the spm ading of radioactive contamin-ation. Other protective equipent such as filter type b.; Phing apparatus vill also be avn11able.
(7) in the cat,e of maintenance jubs performed under ndiation exposure conditions which limit the vorhing period, special check lists and procedures vill be developed and the perre tnel involved vill be thoroughly briefed prior to entry inte the work area.
i (8) The air lock vill be used for personnel access to the. contain-ment vessel at all times, except when loading a complete new com or when the main coolant is less than 150 psig and 3000 F respectively, sna the reactor is suberitical ty at least 5%
under ambient temperature conditim. Whenever the riir lock
.s opened, the scram circuit breaker vill be locked open and no nactivity insertions such as additions of fuel or amoval 1
of boron er control rods will is made.
(9) The equipment access opening vill be used only when the main coolant is less than 150 psig and 300 F, nspectively, and the reactor is suberitical by at least 5% under ambient tes.peratun s
condition.. Whenever the equipsent access door is unbolted, the sens cirt:uit breaker vill be locked open and no reactivity .
insertions such as additions of fuel or nuoval of borm or conti-ol' rods will be We. It-is only. intended to use the equip-7/27/61
_..m_ a - __.____.,_.m_u. _
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ment access opening for the pasange of spent fuel shipping cask.
(1) Refueling and control rod replacements (1) Iteme (1), (2), (3), and (4) and 17 (a) ~ Initial core loading, vill be cceplied with during refueling.
(2) In the case of the removal of an individual control rod, the four adjacent fuel assemblies vill be removed firot.
(3) A minimum of two source range flux channels and two inter-mediate range flux channels vill be in service during these operations.
(4) Radiation monitoring equipment and alarma vill be in service during these operations. ,
(j) Emergencies:
(1) in case of high radioactivity level in the containment vessel indicated by either the air particle detector alam or- <
radioactive gas detector alarm, the reactor will be shut down.
The purge system vill be put into operation if the stack discharge activity does not exceed 2 x 10'I ue/ce.
If this action does not reduce the radioactivity level, a reactor cooldown vill be initiated.
(2) If the stack ndioactivity alarm is actuated, the yardous possible sources of radioactivity vill be shut off sequer.-
tially. If this action does not Maedy the conditicn, the reactor vill be shut down or cooled down as the case may be.
(3) If the steam generator blowdon radioactivity alam is s ctuated, the blowdown vill be discontinued inueediately and the reactor v1'.1 be shut down.
T/27/61 I
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(h) If the radioactive alarm in the liquid effluent line to the river is actuated, the valve in this line vill be closed immediately.
(5) Inesseofhighcoolantradioactivityabove20ue/cc,the 3 purificatica system flow vill be increased and if this does not help to reduce the radioactivity in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the reactor vill be shut down.
(6) Operating instructions vill be pa pared for emergencies such as loss of coolent, loss of coolant flow, control rod anl-function, loes of steam 1 cad, failure of reactor control circuit, and uncon+, rolled heat extraction and these instruc-tions vill enumers.te the automatic acticca tint should take piace and also list the inanediate. manual actions that the opentor should tako. In all of these emergencies, if the automatic control does no+, shut down the reactor, the oper-ator is instructed to shut the reactor down by either manual scram or by running in the centrol rods.
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