ML20148G408

Forwards Initial Eval of 8 Complete SEP Topics & Response to Request for Clarification of SEP Documentation Procedures Made at 780531 Meeting.Requests Response Re Facts Used to Define Plants by NRC Staff.(See ANO 7811130337.)

ML20148G408
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	08/17/1978
From:	Eisenhut D Office of Nuclear Reactor Regulation
To:	Counsil W CONNECTICUT YANKEE ATOMIC POWER CO.
References
TASK-03-10.C, TASK-04-01.A, TASK-04-03, TASK-05-09, TASK-06-07.A2, TASK-06-07.D, TASK-07-01.B, TASK-17, TASK-RR NUDOCS 7811130064
Download: ML20148G408 (2)

Text

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UNITED STATES y" s ., NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555

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COMMONWEALTH EDISON COMPANY DOCKET NO. 50-237 DRESDEN STATION UNIT NO. 2 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 39 License No. OPR .9

1. The Nuclear Regulatory Comission (the Comission) has found that:

A. The applications for amendment by the Commonwealth Edison Company (the licensee) dated September 10, 1974 and May 17, 1976, as sup-plemented March 21, 1977 and March 13, 1978, comply with the stand-ards 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; 3

B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C. There is reasonable assurance (i) that the activities 3uthorized 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 tne 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. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amend-ment, and paragraph 3.B of Provisional Operating License No. OPR-19 is hereby amended to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 39, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

7 811 b O ()()G 4

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3. This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION ha- 2 ode ppolito, Chief Thomas A.

Operating Reactors Branch #3 '

Division of Operating Reactors

Attachment:

I Changes to the Technical Specifications l Date of Issuance: November 13, 1978 4

ATTACHMENT TO LICENSE AMENDMENT NO. 39 PROVISIONAL OPERATING LICENSE NO. DPR-19 DOCKET N0. 50-237 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

Remove Insert 88 88 90 90 91 f 92 92 93 93

, u -,--r--- - - - . . e-, + .._-- , . - - , .., - , ,.--,,

4.6 SUltVEILLANCE REQUlilEMENT 3.6 LIMITING CONDITION FOR OPEIIATION Pressurization Temperature D. Pressurization Temperature B.

1. Reactor vessel shell temperature and  !

1. The reactor vessel shall be vented reactor coolant pressure shall be per-and power operation shall not be manently recorded at 15 minute intervals conducted unless the reactor vessel whenever the shell temperature is below temperature is equal to or greater 220*F and the reactor vessel is not vented.

' than that showr in Curve C of Figure 3.6.1. Operation for hydro- 2. When the reactor vessel head bolting studs static or leakage tests, during are tightened or loosened the reactor ves-

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heatup or cooldown, and with the sel shell teinperature immediately below core critical shall be conducted th head flange shall be permanently only when vessel temperature is equal recorded.

to or above that shown in the appro_

priate curve of Fig. 3.6.1. Figure 3. Neutron flux monitors and samples shall 3.6.1 is effective through 6 effectivo be installed in th reactor vessel a liaeent full power years. At least six months t the vessel wall t the core midplane l f prior to 6 effective full power years level. The monitor and sample program i l new curves will be submitted. shall as a minimum conform to AST.1 \

E 185. The monitors and samples shall be

2. The reactor vessel head bolting studs mm ved and tested during the tidt-d ivfuehng shall not be under tension unless the ""I"E" I" "l# *i"""l""Y *% lh" temperature of the vessel shell " " " "U " "I "" " "' I "k E ""' d "' "l ""

immediately below the vessel flange is >100 F. II"Xl).iat re used to determine the NDTT for 1 agure 4.G.1.

C. Coolant Chemistry C. Coolant Chemistry

1. The reactor coolant system radioactivity 1. a. A sample of reactor coolant shall be taken at least every concentration in water shall not exceed 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> and analyzed for 20 microcuries of total iodine per ml radio-activity, of water

b. Isotopic analysis of a sample of reactor coolant shall be made at least once per month.

Amendment No. 39 88 L_ - - _ __

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1 3.6 LIMITING CONDITION FOR OPERATIOM 4.6 SURVEILLANCE REQUIREMENT .

an orderly shutdcwn shall be initiated and the reactor shall be in a Cold Shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2. The primary containment sump sampling 2. The primary containment sump sampling system and an air sampling system shall and air sampling system operability be operable during power operation. If will be observed daily as pa rt of either a sump water sample or a contain- 4.6.D.2.

ment air sample cannot be obtained for any reason, reactor operation is permissible only during the succeeding seven days unless the system is made operable during this period.

E. Safety and Relief Valves E. Safety and Relief Valves

1. During reactor power operating conditions A minimum of 1/2 of all safety valves shall be bench checked or replaced with and whenever the reactor coolant pressure a bench checked valve each refueling is greater than 90 psig and temperature outages. The popping point of the greater than 320 f, all eight of the safety valves shall be set as follows:

safety valves shall be operable. The Number of Valves Set Point (Psig solenoid activated pressure valves shall be operable as required by Specification 1 1125*

2 1240 3.5.D. 2 17 0 2 l' n

?. 1;:-30 The allowable set point error for each valve is l'A 5 All relief valves shall be checked for i set pressure each re fueling outage. The g set pressures shall be:

A 2. If Specification 3.6.E.1 is not met, an Number of Valves Set Point (Psic 2 orderly shutdown shall be initiated and 1126*

p the reactor coolant pressure and temper- g 33 4 ature shall be 190 psig and 2320 F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. (

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• Target Rock combination safety / relief valve 90

Minimum Temperature Requirements per Appendix G of 10 CFR 50 .;

1200 g E

1100 ,

1. c CURVE A (INSERVICE PRESSURE TESTS-1000 f

! [ SECTION XI)

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/ / I I I CURVE B (HE ATUP - COOLDOWN) g 900

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g f CURVE C (CRITICAL CORE OPERATION)

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O 800 j

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4 g

> f A /

$ 700 #

s2 11 E ll

600 g d

g 3 ll 500 ;j h

] MINIMUM BOLTING TEMPERATURE = 100 0F

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MINIMUM OPER ATING TEMPERATURE = 149 F

$ 400 ll y ll RTNDT = 40 F b ll Kg = PER SECTION G-2110 OF APPENDIX G 300 OF THE SUMMER 1973 ADDENDA TO m

qI [ SECTION lli OF THE ASME CODF I'l 200 100 i

0 -t 100 200 300 i

TEMPERATURE ( F) l Fig. 3.6.1 Amendment No. 39

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rate continued over n 450*F coolant temperr.-

Bases: ture range. _;

A. Thermal Limitations- The reactor vessel The uncontrolled cooldown rate of 240*E was de.wign' specification requires that the reactor based on the maximum expected transient over vessel be designed fw a maximum heatup and cooldown rate of the contained fluid (water) of the lifetime of the reactor vessel. This maxi-Inn'F per hour averaged over a period of one mum expected transient is the injection of cold water into the vessel by the high pressure hour. This rate has been chosen based on coolant injection subsystem. This tr:mstent part experience with operating power plants. was considered in the design of the:preesure The addociated time perists for heatup and vessel and five such cycles were considered conidown cycles when the luu*F per hour rate to limiting provides for efficient, but safe, in the design. Detailed stress analyses were conducted to assure that the injection of cold plant operation. water into the vessel by the ItPCI wouhl not The reactor vessel manufacturer has designed exceed ASME stress code limitations.

the ressel to the above temperature criterion.

In the course of completing the design, the B. Specification 3.6.A.4 increases margin of manufacturer performed detailed stress safety for thermal-hydraulic stability and analvsis. This analysis inchides more severe startup of recirculation pump.

therInal conditions than those which would be encountered during normal heating and cooling ope ratio ns. Pressurization Temperature - The reactor coolant system is a primary barrier against the release specific analyses were made based on a heat- of fission products to the environs. In order to ing and cooling rate of 100*F/ hour applied provide assurance that this barrier is mainhined 3

continuously over a temperature range of 100*F at a high degree of integrity, restrictions haa to 550*F. Ilecause of the slow temperature- been placed on the operating conditions to which time responFe of the massive Ilanges relative it can be subjected. These restrictions on to the :uljacent head and shell sections, cal- inservice hydrostatic testing, on heatup and culated temperatures obtained were 500*F cooldown, and on critical core operation shown (shell) and :50*F (flange) (140*F differedtlat). in Figure 3.6.1, were established to be in con-Iloth asial and radial thermal stresses were formance with Appendix G to 10 CFR 50.

I considered to act concurrently with full pri-marv loadings. Calculated stresses ner In evaluating the adequacy of ferritic steels wittun .\SME Iloiler and Pressure Vessel Code Sa302B it is necessary that the following be Section H1 stress intensity and fatigue limits established:

even at the flange area where maximum stress oCeurs. a) The reference nil-ductility temperature (RTNDT) for all vessel and adjoining The llange metal tempertture differential of materials, I to F occurred as a result of sluggish temper-ature rerponse amt the fact that the heating i

Amendment No. 39 i 92 l

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1-b) the relationship between RTNDT and ferrectic steels.- Provision'has and integrated neutron flux (fluence, been made for the modification of at energies > 1'Mov), and these curves to account for the c) the fluence at the location of a change in RTNDT as a result of postulated flow. neutron embrittlement.

The initial RTHDT of the main closure flange, the shell and head materials connecting to C. . Coolant Chemictrg - A radioactivity these flanges, and connecting welds is 100F. concentration limit of 20 r ci/ml However, the vertical electroslag welds which total iodine can be reached if terminate immediately belo.w the vessel flange the gaseous effluents are near have an RTNDT of 400F. Reference Appendix F the limit as set forth in to the FSAR. The closure flanges and con- Specification 3.8.C.1 or there is.

necting shell materials are not subject.to a failure or a prolonged shutdown any appreciable neutron radiation exposure, of the cleanup domineralizer. In nor are the vertical electroslag seams. The the event of a steam line rupture, flange area is moderately stressed by outside the drywell, the resultant tensioning the head bolts. Therefore, as radiological dose at the site is indicated in curves (a) and' (b) of Figure boundary would be about 10 rem to 3.6.1, the minimum temperature of the vessel the thyroid. This dose was cal-shell immediately below the vessel flange culated on the basis of a total is establbhed as 1000F below a pressure of iodine activity limit of 20 sci /ml, i 400 psig. (400F + 60oF, where 400F is the meteorology corresponding RT NDT of the electroslag weld and 600F is a conservatism required by the ASME Code).

Above approximately 400 psig pressure, the stresses associated with pressurization are more limiting than.the bolting stresses, a fact that is reflected in the non-linear

portion of curves (a) and (b). Curve (c),

which defines the temperature limitations for critical core operation, was established per Section IV 2.c. of Appendix G of 10CFR50.

j Each of the curves, (a), (b) and (c) define temperature limitations for unirradicated  ;

Amendment No. 39