ML19246C470

Forwards Des for Comment.Requests Response by 790807.W/o Encl

ML19246C470
Person / Time
Site:	New England Power
Issue date:	05/31/1979
From:	Ballard R Office of Nuclear Reactor Regulation
To:	ENERGY, DEPT. OF
References
NUDOCS 7907250207
Download: ML19246C470 (2)

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I w( cq/ WASHING TON, D. C. 20555 M.Y 31 1979 Docket '.os. STN 50-568 and STN 50-569 U. S. Department of Energy Washington, D. C. 20455 Gentlemen: I am forwarding for your review and comment the draft environmental impact documentation identified in the enclosure to this letter. The Draf t Environmental Statement was prepared by my staff in accordance with the statement of general policy and procedure on implementation of the National Environmental Policy Act of 1969, as set forth in the Commission's regulations,10 CFR Part 51. The statement has today been sent to the Environmental Protection Agency and notice of its availability is being forwarded to the Office of the Federal Register for publication. Comments will be due within 60 days after publication in the Federal Register of the Environmental Protection Agency's listing notifying the public of issuance of the impact statement. In particular please provide coments on Appendix 0, which deals with uranium resources availability. Sincerely, A 52 dC';d/'$o/i d 6nald L. Ballard, Chief Environmental Projects Branch 1 Division of Site Safety and Environmental Analysis

Enclosure:

NEP 1 & 2 DES 56Sr?56 7007250267 -2%

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~ An upper bound envelope as defined by the nor=alized peaking factor axial dependence of Figures 3.2-3, a, & b has been determined to be consistencwiththel technical specifications on power distribu:: ion control as givec in Section 3.2. When aa ? measure = ant is taken, both experi: ental error and =anufacturing 9 tolerance must be allowed for. Five pe: cent is the appropriate experi= ental uncerta.inty allowance for a full core map taken with the movable incore - detector fim =apping systen and three percent is the appropriate allowance for manufact d g. tolerance.. In the-spec f Hed linf t o'f Y, there is an 8 percent allowance for uncertain-ties which means that no--21 operation of the core is expected to result in JI e g l.55/1.08. The logic behind the larger uncertainty in this case is that (a) nor=21 perturbatio=3 in the radial power shape (e. g., rod misalign-Ji ment) affect M,,in most casas without necessarily. af fecting F, (b) altho, ugh the x q operator has a direct influence on F through cove =ent of rods, nm! < an li:nt 9 it to the desired value, he has ne direct control over and (c) an error in the predictions for radial power shape, which may be detected during startup physics tests can be co=pensated for in T by tighter axial control, y 9 but compensation for F' in less readily available. Vnen a measuresant of g M is taken, c peri =antal error =ust be allowed for and 4". is the appro-priate nilowance for a full core =ap taken with the covable incore detector flux capping r;ystem. ~ Measure =ents of the hot channel factors are required as part of start-up physics tests, at least once each full rated power month of operation, and whenever abnor=al poser distribution conditions require a reduction of core power to a level based on measured hot channel factors. The incore cap taken followlag. initial loading provides confirmation of the basic nuclear P00R ORGM

== B3.?-4 Amendment t;o. 49, linit 3 Amendment t;o. 41, Unit 4

~ Flux Di.ffere=ca (Ac) a:d a ref ere=r.c value which corresponds to the full desiga pouer eq,*14heit=s value of Artal of fset (Arial of f set - 4/fractiosal I povar). The reference value of flux difference varies with power level and bur =up. but expressed as axial offset it varies only vith burnup., Tha t"-b f m i-apacificaticus on power distribution control assure that the l F upper bound anvelope as defined by Figures 3.2-3,a,6 b is not exceeded and xenon distributices are not developed which at a later time, vould cause graater local power pa*4ny even though the flu = diff erence is then W- tha,1 d-i ts spec *" bed by the p =cedure. ~ The target (or refere=cs) value of flux diff erence is deter =ined._a3 follows. At a=y ti== that equilibrit=s ze=en conditio=s have been establish-d, tha in-dicated flu: difference is noted with part length +rcris withdrawn feca the core and with the full le=gth rod control rod bank core th2n 190 steps vi $ drawn (i.e., vor=al rated power operating positica appropriate for the *4 = in lif e. Control rods are usually withdr:nzn farder as burnup proceeds). This value, divided by the fraction of design power at which the core was operating is the design power value of the target flux difference. Values for all other core power 1-vels are obm'ned by cultiplying the design pcver value by the fractional power. Since the indicated equ'i dhrinn value was =sted, no allowm-rns for eznore detector error are necessary mai 4-d',ted deviation of +5~ AI are p

• tted frc= the indicated reference valne.

De--ing pe-icda where exte:mive load following is required, it =ay be impractical to establish the required core conditices for ~, :urdsg the target flu differe=ce eve y rated power month. For this reason, cethods are pa- 'tted by Ite= 6c of j Section 3.2 for updating the target flux differences. Fis;ure E3.2-1 shova a typical constructica of the target flux differe=ce band at BOL and Fi;;nre 33.2-2 shows the typical vn *, tion of the full power value with burcup. Strict cont ci of the flux difference (and red positica) is not as necessary during part power operation. This is bec use xc=on distribution co==cl at part power la not as significant as the cent.ol at full power and aUlovance has been cade in predicting the he2t flux peaking factora for less :=1ct ec=- trol at pa-- pcaer. Strict control of the flux difference is cet possible during cer,f n physics tests or during the required, pericdic excore H ' bra-q q<> c4 uud DJL n-i (tg' +An reference to part-length rods no longer applie i . rods are r r.oved.rca the reactor. I J ay B3.2-6 Amendment No. 49, Unit 3 .- N t "n n, t. B. Pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended. C. Pursuant to the Act and 10 CFR Parts 30, 40 and 70 to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required. D. Pursuant to the Act and 10 CFR Part 30 to receive, possess, and use at any time 100 millicuries each of any byproduct material without restriction to chemical or physical form, for sample analysis or instrument calibration; E. Pursuant to the Act and 10 CFR Parts 40 and 70 to receive, possess, and use at any time 100 milligrams each of any source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration; F. Pursuant to the Act and 10 CFR Parts 30 and 70 to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of Turkey Point Unit Nos. 3 and 4. 3. This license shall be deemed to contain and is subject to the conditions specified in the following Commission Regulations in 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 4L;.41 of 10 CFR Part 40, Section 50.54 and 50.59 of 10 CFR Part 50 and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below: A. Maximum Power Level The reactor shall not be made critical until the tests described in the applicant's letter of April 3,1973, have been satisfactorily completed. Thereafter, the applicant is authorized to operate the facility at reactor core power levels not in excess of 2200 megawatts thermal. t yw'~~h}. 563262 g gggt B. Technical Specifications The Technical Specifications contained in Appendices A and B as revised through Amendment No. 41 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. C. This license is subject to the following conditions for the protection of the environment: (1) The applicant shall pursue evaluations of alternatives to the proposed cooling channel systen during construction, interim operation, and evaluation of the channel system. These evaluations shall include at least the following: (a) Study of availability of groundwater or other alternative sources of surface water to use in the cooling system. (b) Study of applicability of mechanical cooling cevices, including powered spray modules and cooling towers. (c) Study of marine environmental impacts of once-through cooling alternatives (described in Section X of the AEC Final Environmental Statement on Turkey Point Units 3 and 4, July 1972). (2) The applicant shall take appropriate corrective action on any adverse effects determined.ls a result of monitoring and study programs. To the fullest extent practicable, the applicant shall utilize results of study programs in improving and modifying the operation of the facility and its cooling system so as to achieve a ainimal adverse environmental impact. D. Steam Generator Operation (1) Af ter equivalent operation in Cycle 6 of six months from June 1,1979, Turkey Point Unit 4 shall be brought to the cold shutdown condition and the steam generators shall be inspected pniess: (1) an inspection of the steam generators EGSr?G3 is performed within this six month period as a result of the requirements in 2, 3 and 4 below, or (2) an acceptable analysis of the susceptibility for stress corrosion cracking of tubing is submitted to explicitly justify continued operation of Unit No. 4 beyond the authorized six equivalent months of operation. Any analysis justifying continued operation must be submitted at least 45 days prior to the expiration date of the authorized six equivalent months of operation. For the purpose of this requirement, equivalent operation is defined as operation with the reactor coolant at a temperature greater than 350 F. Nuclear Regulatory Commission (NRC) approval shall be obtained before resuming power operation following this inspection. (2) Reactor coolant to secondary leakage through the steam generator tubes shall be limited to 0.3 gpm per steam generator,. With a steam generator tube leakage greater than this limit, the reactor shall be brought to the cold shutdown condition within 24 hours. The leaking tube (s) shall be evaluated and plugged prior to resuming power operation. (3) The concentration of radiciodine in the reactor coolant shall be limited to 1.0 microcurie / gram during normal operation and to 30 microcuries/ gram during power transients. (4) Reactor eperation shall be terminated and NRC approval shall be obtati.ad prior to resuming operation if primary to secondary leakage attributable to the denting phenomena is detected in 2 or more tubes during any 20 day period. (5) The Metal Impact Monitoring Systen (MIMS) shall be contained in operation with the capability of detecting losse objects. If the MIMS is out of service in other than cold shutdown or refueling mode of operation, this fact shall be reported to the NRC. Any abnormal indications from the MIMS shall also be reported to the NRC by telephone by the next working day and by a written evaluation within two weeks. (6) Following each startup from below 350*F, core barrel movement shall be evaluated using neutron noise tecnniques. 563264 E. The licensee shall maintain in effect and fully implement all provisions of the Commission-appro"ed physical security plan, including amendments and changes made pursuant to the authority of 10 CFR 50.54(p). The approved security plan documents, withheld from public disclosure pursuant to 10 CFR 2.790(d), collectively titled " Turkey Point Plant Unit Nos. 3 and 4 Physical Security Plan", dated October 18, 1978, as supplemented Feb rua ry 20, 1979". F. Fire Protection The licensee may proceed with and is required to provide a schedule for and to complete the modifications identified in Paragraphs 3.1.1 through 3.1-19 of the NRC's Fire Protection Saf ety Evaluation, dated May 21, 1979 for the facility. These modifications are to be completed prior to December 1980. If any modifications cannot be completed on schedule the licensee shall submit a report explaining the circumstances together with a revised schedule. In addition, the licensee shall submit the additional information identified in Sections 3.1 and 3.2 of the related Safety Evaluation in accordance with the schedule contained therein. In the event these dates for submittal cannot be met, the licensee shall submit a report, explaining the circumstances, together with a revised schedule. The licensee is required to develop and implement the administrative controls which are consistent with the licensee's letters of August 28 and November 7, 1978 within three months from the date of this amendment. 4. This license is effective as of the date of issuance, anc chall expire at midnight April 27, 2007. FOR THE ATOMIC ENERGY COMMISSION Original Signed By A. Giambusso, Deputy Director for Reactor Projects Directorate of Licensing Attachments: Appendix A - Technical Specifications Appendix B - Environmental Technical Specifications e v e r, -. u G C 4 b t> Date of Iss iance: April 10, 1973}}