ML17244A365
| ML17244A365 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 01/30/1979 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17244A366 | List: |
| References | |
| NUDOCS 7902210017 | |
| Download: ML17244A365 (18) | |
Text
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k1tei UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 ROCHESTER GAS AND ELECTRIC CORPORATION DOCKET NO. 50-244 R.
E.
GINNA NUCLEAR POWER PLANT AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No.
23 License No.
DPR-18 l.
The Nuclear Regulatory Collnlission (the Collnlission) has found that:
A.
B.
C.
D.
E.
The application for amendment by Rochester Gas and Electric Corporation (the licensee) dated April 30, 1976, complies with the standards and requirements of the Atomic Energy Act of
- 1954, as amended (the Act), and the Commission's rules and requlations set forth in 10 CFR Chapter I; i
The facility will operate in conformity with the aoplication, the orovisions of the Act, and the rules and requlations of the Commission; 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 Comission's requla-tlons~
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 The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
p V V
~
y ll
2.
Accordingly, the license is amended by changes to the Technical Soecifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Provisional Ooerating License No.
DPR-18 is hereby amended to read as follows:
(2)
Technical S ecifications The Technical Specifications contained in Appendix A as revised through Amendment No.
23, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COhNISSION
Attachment:
Changes to the Technical Specifications Dennis L. Ziemann, hief Operating Reactors Branch
>2 Division of Operating Reactors Date of Issuance:
January 30, 1979
ATTACHMENT TO LICENSE AMENDMENT'NO. 23 PROYISIONAL OPERATING LICENSE NO.
DPR-18 DOCKET NO. 50-244 Remove the following pages of the Appendix "A" Technical Specifications and insert the enclosed pages.
The revised pages contain the captioned amendment number and vertical lines indicating the areas of change.
REMOVE 3.1-2 3.1-3 3.1-20 4.1-6 4.1-8 4.1-9 4.1-10 INSERT 3.1-2 3.1-3*
3.1-20 4.1-6 4
1 8**
4.1-9 4.1-10 he revised portion now appears on Page 3.1-2.
This page is included for administrative purposes only.
- This page is included merely for the purposes of adding the appropriate Table number.
3.1.1.1 d.
At least one reactor coolant pump shall be in operation for a planned transition from one Reactor operating Node to another involving an increase in the boron concentration of the reactor
- coolant, except for emergency boration.
3.1.1.2 Steam Generator a.
One steam generator shall be capable of performing its heat transfer function whenever the average coolant temperature is above 350'F.
b.
The temperature difference across the tube sheet shall not exceed 100'F.
3.1.1.3 Safet Val ves a.
At least one pressurizer code safety valve shall be operable whenever the reactor head is on the vessel.
b.
Both pressurizer code safety valves shall be operable whenever the reactor is critical.
Basis:
When the boron concentration of the reactor coolant system is to be reduced the process must be uniform to prevent sudden reactivity changes in the reactor.
Mixing of the reactor coolant will be sufficient to prevent a sudden increase in reactivity if at least one reactor coolant pump or one residual heat removal pump is running while the change is taking place.
The residual heat removal pump will circulate the primary system volume in approximately one half hour.
The pressurizer is of no concern because of the
')ow pressurizer volume and because the pressurizer boron concentration will be higher than that of the rest of the reactor coolant.
When the boron concentration of the reactor coolant system is to be increased, the process must be uniform to prevent sudden reactivity increases in the reactor during subsequent startup of the reactor coolant pumps.
Mixing of the reactor coolant will be sufficient to maintain a uniform boron concentration if at least one reactor coolant pump is running while the change is taking place.
Emergency boration without a reactor coolant pump in operation is not prohibited by this specification.
- 3. 1-2 Change No.
72 Amendment No.
The specification requires that a sufficient number of reactor coolant pumps be operating to provide core cooling.
The flow provided in each case will keep DNB well above 1.30 as discussed in FSAR Section 14.1.6.
Therefore, cladding damage and release of fission products to the reactor coolant will not occur.
Heat transfer analyses(1)show that reactor heat
'I equivalent to 130 HHT (8.5I) can be removed with natural circulation only;
- hence, the specified upper limit of ll rated power without operating pumps provides a substantial safety factor.
Each of the pressurizer code safety valves is designed to relieve 288,000 lbs.
per hr. of saturated steam at the valve set point.
Below 350'F and 350 psig in the reactor coolant system, the residual heat removal system can remove decay heat and thereby control system temperature and pressure.
If no residual heat were removed by any of the means available the amount of steam which could be generated at safety valve relief pressure would be less than half the valves'apacity.
One valve therefore provides adequate defense against overpressurization.
- 3. 1-3 Change No. gg Amendment No. 23
3.1. 4 Maximum Coolant Activit
~Elf'.1.4.1 3.1.4. 2 Whenever the reactor is critical or the reactor coolant temperature is greater than 500'F:
a.
The total specific activity of the reactor coolant shall not exceed 84/E~Ci/gm, where E is the average beta and gamma energies per disintegration in Mev.
b.
The I-131 equivalent of the iodine activity in the reactor coolant shall not exceed 3.0~Ci/gm.
c.
The I-131 equivalent of the iodine activity on the secondary P
side of a steam generator shall not exceed 0.1~Ci/gm.
If any one of the activity limits in Specification 3.1,4.1, (a.
and b.) are exceeded due to a power transient:
a.
The activity shall be returned to within specification within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, or b.
The plant shall be brought to a hot shutdown condition within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Basis:
The total activity and iodine activity limits for the primary system correspond to operation with the plant design basis of lX fuel defects.
Radiation shielding and the radioactive waste disposal systems were f
3.1-20 Amendment No. 5, 23
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TABLE 4.1-1 (CONTINUED)
Channel Descri tion 10.
Rod Position Bank Counters ll.
Steam Generator Level S(1,2)
N.A.
Check Calibrate Test N.A.
Remarks 1)
Fach s ix inches of rod motion when data logger is out of service
- 2) With analog rod position 12>
Charging Flow 13.
Residual lleat Removal Pump Flow 14.
Boric Acid Tank Level N.A.
8.A.
D R
N.A.
N.A.
N.A.
Bubbler-tube rodded weekly 15.
Refueling Water Storage Tank Level N.A.
N.A.
16.
Volume Control Tank Level 17.
Reactor Containment Pressure 18.
Radiation Monitoring System 19.
Boric Acid Control 20.
Containment Drain Sump Level N.A.
D N.A.
N.A R
N,A.
M (1)
- 1) Isolation Valve signal M
N.A.
N.A.
21.
Valve Temperature Interlocks N.A.
N.A.
22.
Pump-Valve Interlock N.A.
N.A.
23.
Turbine Trip Set-Point N.A.
M(Q
- 1) Block trip 24.
Accumulator Level and Pressure S
N.A.
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Table 4.1-2 Teat pc'
~
S walla ~
Sec:i o.".
1.
P.eactor Cool" nt S" n.pic s W)onthiy <<)
.'vfont'r.ly {6)
V4)'ccl-ly (61 3 times/week and at least every third day Gross Radioactivity 3 times/ivcc):ly Conccnt,ration and at least (beta-ganesa) eVCry third aay (l ) {4)
I'adio-chc)T)ical {2)(4)
P Detcrn~i)lation (2)
Tritium Conccnt "ation C,"ra'oridc and Fluoride P~eeen 5 times/week and at least every second day except when below 250'F G'0 s s Rad) o)o')inc Concento,r ltion eccl;ly (3) {6) 2,.
Reactor Cool-nt Boron Boron Con C C)) Ll'ca't)on Vfcekly Refuell')" Vacate r
'.Ora" C Tc)));i Vo O'Lc)'allop ~ c Bo Ton con" cnt ration Vl'cckly
) ~
Do)'1 c Ac )d fc'k I3ol'Dn concc>>4 ration TN'ic c/v:c e!~
5.
Co>>tro) Rods Rod'drop time.
of a ful) )c))g.'h rods Each rcfuc'in~
sl) u td oiv))
Pull Lcn)-,th Control
'Rod
'arti~i move )lent of all rods Every 2 ~ve l;s {6) 7.
Prcssuri cr Safety Valves S.
Vaain Stcam Safety Valves Set point Set po)nt Each Refueling shutdoinl Each.
Refueling'ilutdov:n
}o 9.
Containmcnt Isolation Trlp Function ng Each Refuelir>> sl)utc!o'v::
Rcfucl)ng S) =-'.";..
lntcl locks I unCtiOni)lc Prior to refueling 9, ~, "
operations Amendment No. 5, 23 4.1-8
. ~
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Tabl e 4.1-2 (Continued)
Test
~Fre uenc FSAR Sect>on Re fn, ence 11.
Service Water. Systemi Functioning Each Refueling shutdown Fire Protection Pump and Power Supply
, 12;
,13.
l1onthly Functioning Spray Additive Tank Accumulator HaOH concentration Y>onthly 14.
Boron concentration Bi-monthly Daily Daily
.16.
Diesel Fu 1 Supply Fuel inventory 15.
Primary Systen Leakage Evaluate 9.5.5 8.2.3 1.7; Spent Fuel Pit 18.
Secondary Cool ant Samples
'9.
Circulating Water Flood Protection Equipment Boron concentration Gross activity Calibrate llonthly 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (5)(6)
Each Refueling Shutdown 9.5.5 hot s:
'1)
A gross radioactivity analysis shall consist ot'he quantitative measurement of the total radioactivity of the primary coolant in units of pCi/gm.
The total primary coolant activity shall be the sum of the degassed beta-gamma activity and the total of all identified gaseous activities 15 minutes after the primary system is. sampled.
Whenever the gross radioacticity concentration exceeds 10,". of the limit specified in the Specif:cation 3.1.4.1.a or increa"es y
Anendrpent Ho. lf, 23 4.1-9
/
P
10~Ci/gm from the previous measured level, the sampling frequency shall be increased to a minimum of once/day until a steady activity level is established (2)
A radiochemical analysis shall consist of the quantitative measure ment of the activity for each radionuclide which is identified in the primary coolant 15 minutes after the primary system is sampled The activities for the individual isotopes shall be used in the determination of E.
A radiochemical analysis and calculation of E and iodine isoto ic activity shall be performed if the measured gross activity changes by more than 10~Ci/gm from the previous measured levels (3)
In addition to the weekly measurement, the radioiodine concentration shall be determined if the measured gross radioactivity concentration changes by more )han 10~Ci/gm from the previous measured level (4)
Iodine isotopic activities shall be weighted to give equivalent I-131 activity (5)
An isotopic analysis for DOSE E(UIVALENT I-131 concentration is required at leas't once every 31 days whenever the gross activity determination indicates iodine concentration greater than 10% of the allowable limit but only once per 6 months whenever the gross activity determination indicates iodine concentration below 10K of the allowable limit (6)
Not required during a cold or refueling shutdowns (7)
During a cold or refueling shutdown, primary coolant Gross Radio activity will be determined weekly 4.1-10 Amendment No. 5, 23