ML20067B315
| ML20067B315 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 02/10/1994 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Metropolitan Edison Co, Jersey Central Power & Light Co, Pennsylvania Electric Co, GPU Nuclear Corp |
| Shared Package | |
| ML20067B317 | List: |
| References | |
| DPR-50-A-182 NUDOCS 9402240150 | |
| Download: ML20067B315 (9) | |
Text
uq j
n Mo g$p t
UNITED STATES 0
NUCLEAR REGULATORY COMMISSION f
WASHINGTON. D.C. 20666-0001
\\...../
METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY GPU NUCLEAR CORPORATION DOCKET NO. 50-289 l
THREE MILE ISLAND NUCLEAR STATION. UNIT NO 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 182 License No. DPR-50 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The applications for amendment by GPU Nuclear Corporation, et al.
(the licensee), dated May 26 and December 2, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
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.
9402240150 940210 PDR ADDCK 05000289 P
. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and parag aph 2.c.(2) of Facility Operating License No.
DPR-50 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.182, are hereby incorporated in the license.
GPU Nuclear Corporation shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
\\
r
_[
- l
~\\f },g'&f Joh F. Stolz, Directer Pr ect Directorate I--4 Di' vision of Reactor Projects - I/II Office of Nuclear Reactor Regulation N
Attachment:
Changes to the Technical Specifications Date
' suance:
February 10, 1994 i
l 1
ATTACHMENT TO LICENSE AMENDMENT N0.182 FACILITY OPERATING LICENSE NO. DPR-50 DOCKET NO. 50-289 i
Replace the following pages of the Appendix A Technical Specifications with the attached pages.
The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
Remove Insert 11 11 3-40f 3-40f 3-59 3-59 3-105 3-105 3-106 3-106 f
4-7a 4-7a i
i l
l j
l
\\
TABLE OF CONTENTS Section Pace 2
SAFETY L'MITS AND LIMITING SAFETY SYSTEM SETTINGS 2-1 2.1 Safety Limits. Reactor Core 2-1 2.2 Safety Limits. Reactor System Pressure 2-4 2.3 Limitina Safety System Settinas. Protection Instrumentation 2-5 3
LIMITING CONDITIONji FOR OPERATION 3-1 3.0 General Action Reauirements 3-1 3.1 Reactor Coolant System 3-la 3.1.1 Operational Components 3-la 3.1.2 Pressurization, Heatup and Cooldown Limitations 3-3 3.1.3 Minimum Conditions for Criticality 3-6 3.1.4 Reactor Coolant System Activity 3-8 3.1.5 Chemistry 3-10 3.1.6 Leakage 3-12 3.1.7 Moderator Temperature Coefficient of Reactivity 3-16 3.1.8 Single Loop Restrictions 3-17 3.1.9 Low Power Physics Testing Restrictions 3-18 3.1.10 Control Rod Operation 3-18a 3.1.11 Reactor Internal Vent Valves 3-18c j
3.1.12 Pressurizer Power Operated Relief Valve (POR'I) j and Block Valve 3-18d 3.1.13 Reactor Coolant System Vents 3-18f 3.2 Makeup and Purification & Chemical Addition Systems 3-19 i
3.3 Emeroency Core Coolina. Reactor Buildina Emoraency Coolina and Reactor Buildina Sorav Systems 3-21 3.4 Decay Heat Removal Capability 3-25 3.4.1 Reactor Coolant System Temperature Greater than 250*F 3-25 1
3.4.2 Reactor Coolant System Temperature 250*F or Less 3-26 3.5 Instrumentation Systems 3-27 3.5.1 Operational Safety Instrumentation 3-27 3.5.2 Control Rod Group and Power Distribution Limits 3-33 3.5.3 Engineered Safeguards Protection System Actuation Setpoints3-37 3.5.4 Incore Instrumentation 3-38 3.5.5 Accident Monitoring Instrumentation 3-40a 3.5.6 Deleted 3-40f 3.6 Reactor Buildina 3-41 3.7
. Unit Electrical Power System 3-42 3.8 fuel Loadina and Refuelina 3-44 3.9 Deleted 3-46 3.10 Miscellaneous Radioactive Materials Sources 3-46 3.11 Handlina of Irradiated Fuel 3-55 3.12 Reactor Buildina Polar Crane 3-57 3.13 Secondary System Activity 3-58 3.14 Flood 3-59 3.14.1 Periodic Inspection of the Dikes Around THI 3-59 3.14.2 Flood Condition for Placing the Unit in Hot Standby 3-60 3.15 Air Treatment Systems 3-61 3.15.1 Emergency Control Room Air Treatment System 3-61 3.15.2 Re. actor Building Purge Air Treatment System 3-62a 3.15.3 Auxiliary and fuel Handling Building Air Treatment System 3-62c 3.15.4 Fuel Handling Building ESF Air Treatment System 3-62e j
~
ii Amendment No. M, 72, 7E, 97, S, JJS, J22, JM, JM,167,182 j
1
L 3.5.6 Deleted t
P i
i i
l 1
i Amendment No. 182 3-40f i
-\\
i
3.14 f.L@D 3.14.1 PERIODIC INSPECTION OF THE DIKES AROUND TMI Aeolicability Applies to inspection of the dikes surrounding the site.
Obiective To specify the minimum frequency for inspection of the dikes and to define the flood stage after which the dikes will be inspected.
Soecification 3.14.1.1 The dikes shall be inspected at least once every six months and after the river has returned to normal, following the condition defined below:
- a. The level of the Susquehanna River exceeds flood stage; flood stage is defined as elevation 307 feet at the Susquehanna River Gage at Harrisburg.
Bases The earth dikes are compacted to provide a stable impervious embankment that protects the site from inundation during the design flood of 1,100,000 cfs.
The rip-rap, provided to protect the dikes from wave action and the flow of the river, continues downward into natural ground for a minimum depth of two feet to prevent undermining of the dike (References 1 and 2).
Periodic inspection, and inspection of the dikes and rip-rap after the river has returned to normal from flood stage, will assure proper maintenance of the dikes, thus assuring protection of the site during the design flood.
Refereneg1 (1) UFSAR, Section 2.6.5
" Design of Hydraulic Facilities" (2) UFSAR, Figure 2.6-17
" Typical Dike Section" 4
3-59 Amendment No. 157 182
)
TABLE 3.21-2 (Continued)
TABLE NOTATION
- At all times.
During waste gas holdup system operation.
Operability is not required when discharges are' positively controlled through the closure of WDG-V47 and where RM-A8 (or RM-A4 and RM-A6), FT-149, and FT-150 are operable.
During Fuel Handling Building ESF Air Treatment System Operation.
i
- At all times during containment purging.
6
- At all times when condenser vacuum is established.
ACTION 25 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank may be released to the environment provided that prior to initiating the release:
'l.
At least two independent samples of the tank's contents are analyzed in accordance with Table 4.22-2, Item A, and 2.
At least two technically qualified members of the Unit staff I
independently verify the release rate calculations and verify i
the discharge valve lineup.
3.
The Operations & Maintenance Director, Unit 1, shall approve each release.
Otherwise, suspend release of radioactive effluent via this pathway.
i ACTION 26 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
ACTION 27 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and the initial samples are analyzed for gross activity (gamma scan) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the channel has been declared inoperable.
If RM-A9 is declared inoperable, see also Specification 3.5.1, Table 3-5.1, Item C.3.f.
ACTION 30 1.
With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, a grab sample shall be collected and analyzed for the inoperable gas channel (s) at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
With both channels inoperable, a grab sample shall be collected and analyzed for the inoperable gas channel (s):
(a) at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during degassing operations.
(b) at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during other operations (e.g. Feed and Bleed).
3-105 Amendment No. U, 7S, JH, JU, D7,182
3.22 RADI0 ACTIVE EFFLUENT 3.22.1 LIOUID EFFLUENT 3.22.1.;
CONCENTRATION LIMITING CONDITION FOR OPERATION 3.22.1.1 The concentration of radioactive material released at anytime from the unit to unrestricted areas (see Figure 5-3) shall be limited to ten times the concentrations specified in 10 CFR Part 20.1001-20.2401, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 3 x 10-' uCi/cc total activity.
APPLICABILITY: At all times ACTION:
With the concentration of radioactive material released from the unit to unrestricted areas exceeding the above limits, immediately restore concentration within the above limits.
BASES This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluent from the unit to unrestricted areas will be less than ten times the concentration levels specified in 10 CFR Part 20.1001-20.2401, Appendix B, Table 2.
This I
limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will not result in exposures with (1) the Section II.A design objectives of Appendix I,10 CFR Part 50, to a HEMBER OF THE PUBLIC and (2) the limits of 10 CFR Part 20.1301 to the population. The concentration limit for noble gases is based upon the assumption the Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.
i 1
3-106 Amendment No. 72, 137, JJS, fyg, 180, 182 i
TABLE 4.1-1 (Continued)
CHANNEL DESCRIPTION CHECK IESI CALIBRATE REMARKS 49.
Saturation Margin Monitor S(l)
M(1)
R (1)When T is greater than 525 F.
m 50.
Emergency Feedwater Flow NA M(1)
F (1)When T is greater than 25C*F.
3 m
R Instrumentation E
- 51. Heat Sink Protection System 2
a.
EFW Auto Initiation (1) Includes logic test only.
P Instrument Channels
- 1. Loss of Both Feedwater HA Q(1)
F y"
Pumps
- 2. Loss of All RC Pumps NA Q(1)
R
%3?
- 3. Reactor Building NA Q
F
- y Pressure g
- 4. OTSG Low Level W
Q R
b.
R
~M Pressure O
c.
EFW Control Valve Control System
- 1. OTSG Level Loops W
Q R
R
- 2. Controllers W
NA R
g d.
HSPS Train Actuation Logic NA Q(1)
R
'g 52.
Backup Incore Thermocouple M(1)
NA R
(1)M. hen T is greater than 250"F.
m Display N
- 53. Deleted 54.
RCS Inventory Trending System a.
Level NA NA F
b.
Void Fraction W
NA F
1