ML20151X920
| ML20151X920 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/25/1988 |
| 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 | |
| ML20151X923 | List: |
| References | |
| DPR-50-A-138 NUDOCS 8805040288 | |
| Download: ML20151X920 (8) | |
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UNITED STATES o
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NUCLEAR REGULATORY COMMISSION g
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W ASHING TON, D. C. 20555
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METROPOLITAN EDIS0N COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY GPU NUCLEAR CORPORATION f
DOCKET NO. 50-289 THREE MILE ISLAND NUCLEAR STATION, UNIT NO. 1
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AMENDMENT TO FACILITY OPERATING LICENSE bPR-50 o
1.
The Nuclear Regulatory Comission (the Comission) has found that:
l A.
The application for amendment by GPU Nuclear Corporation, et al.
(thelicensee)datedJanuary 12, 1988 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission'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 Comission; i
C.
There is reasonable assurance (1) 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 regulations; D.
The issuance of this amendment will not be inimical to the ccmon defense and security or to the health and safety of the public; and E.
The issuance of this smendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
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2-2.
Accordingly, the license is amended by changes to the Technical 4
Specificat!ons as indicated in the attachment to this license l
ameadment, and paragraph 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.138 are hereby incorporated in the license. GPU Nuclear Corporation shall operate the facility in accordance with the Technical Specifications.
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3.
This license amendment is effective as of its date of issuance.
l FOR THE NUCLEAR REGULATORY COMMISSION D
hn F. Stolz, Director
?ro' ject Directorate I-4 1
31 vision of Reactor Projects 1/11 Office of Nuclear Reactor Regulation
Attachment:
l Changes to the Technical Specifications Date of Issuance: April 25,1988 l
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ATTACHMENT TO LICENSE AMENDMENT NO.138 FACILITY OPERATING LICENSE NO. OPR-50 DOCKET NO. 50-289 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 i
4-1 4-1 4-2 4-2 i
4-10 4-10 5-5 5-6 5-7 57 l
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l 4.
SURVEILLANCE STANDARDS During Reactor Operational Conditions for which a Limiting Condition for Operation does not require a system / component to be operable, the associated surveillcnce requirements do not have to be performed. Prior to declaring a system / component operable, the associated surveillance requirement must be current.
The above applicability requirements assure the operability of systems / components for all Reactor Operating Conditions when required by the Limiting Conditions for Operation.
4.1 OPERATIONAL SAFETY REVIEW Applicability Applies to items directly related to safety limits and limiting conditions for operation.
Obj ec tiva To specify the minimum frequency and type of surveillance to be applied to unit equipment and conditions.
Specification 4.1.1 The minimum frequency and type of surveillance required for reactor protection system, engineered safety feature protection system, and heat sink protection system instrumentation when the reactor is critical shall be as stated in Table 4.1-1.
4.1.2 Equipment and sampling test shall be performed as detailed in Tables 4.1-2 and 4.1-3.
4.1.3 Each post accident monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the check, test and calibration at the frequencies shown in Table 4.1-4 1
Bases Check TiTTiires such as blown instrument fuses, defective indicators, or faulted amplifiers which result in "upscale" or "downscale" indication can be easily recognized by simple observation of the functioning of an instrument or system. Furthemore, such failures are, in many cases, revealed by alarm or annunciator action. Comparison of output and/or state of independent channels measuring the same variable supplements this type of built-in surveillance.
Based on experience in operation of both conventional and nuclear systems, when the unit is in operation, the minimum checking frequency stated is deemed adequate for reactor system instrumentation.
The 600 ppeb limit in Item 4, Table 4.1-3 is used to meet the requirements of Section 5.4 Under other circumstances the minimum acceptable boron concentration would have been zero ppeb, i
Amendment No. W,9% }D6 M.138
t Calibration Calibration shall be performed to assure the presentation and acquisition of accurate information. The nuclear flux (power range) channels amplifiers shall be checked and calibrated if necessary, every shif t against a hect balance standard. The frequency of heat balance checks will assure that the difference between the out-of-core instrumentation and the heat balance remains less than 4%.
Channels subject only to "drif t" errors induced within the instrumentation itself can tolerate longer intervals between calibrations. Process system instrumentation errors induced by drift can be expected to remain within acceptance tolerances if recalibration is performed at the intervals of each refueling period.
Substantial calibration shifts within a channel (essentially a channel failure) will be revealed during routine checking and testing procedures.
Thus, minimum calibration frequencies set forth are considered acceptable.
Testing On-line testing of reactor protection channels is required monthly on a rotational basis. The rotation scheme is designed to reduce the probability of an undetected failure existing within the system and to minimize the likelihood of the same systematic test errors being introduced into each redundant channel.
The rotation schedule for the reactor protection channels is as follows:
Channels A, B, C & D Before Startup, when shutdown greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Channel A One Week After Startup Channel B Two Weeks Af ter Startup Channel C Three Weeks After Startup Channel D Four Weeks After Startup The reactor protection system instrumentation test cycle is continued with one channel's instrumentation tested each week.
Upon detection of a failure that prevents trip action in a channel, the instrumentation associated with the protection parameter failure will be tested in the remaining channels.
If actuation of a safety channel occurs, assurance will be required that actuation was within the limiting safety system setting.
The protection channels coincidence logic, the control rod drive trip breakers and the regulating control rod power SCRs electronic trips, are trip tested monthly. The trip test checks all logic combinations and is to be performed on a rotational basis.
The logic and breakers of the four protection channels and the regulating control rod power SCRs shall be trip tested prior to startup when the reactor has been shutdown for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Discovery of a failure that prevents trip action requires the testing of the instrumentation associated with the protection parameter failure in the remaining channels.
4-2 Amendment No. J8', }M,138
TABLE 4.1-3 Cont'd.
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Item Check Frequency 4.
Spent Fuel Pool Boron concentration greater Monthly and af ter e,- A eakeup.
Water Sample than or equal to 600 ppeb 5.
Secondary Coolant Isotopic analysis for DOSE At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when h
System Activity EQUIVALENT I-131 concentration reactor coolant system pressure is greater than 300 psig or Tay
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k is greater than 200*F 6.
Boric Acid Mix Tank Boron concentration Twice weekly ***
or Reclaimed Boric Acid Tank C
7.
Deleted co 8.
Deleted 9.
Deleted
- 10. Sodium Hydroxide Tank Concentration Quarterly and af ter each makeup.
- 11. Deleted
- 12. Deleted
- Until the specific activity of the primary coolant system is restored within its limits.
Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since the reactor was last subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.
- Deleted
- The surveillance of either the Boric Acid Mix Tank or the Reclaimed Boric Acid Tank is not necessary when that raspective tank is empty.
5.4 NEW AND SPENT FUEL STORAGE FACILITIES Applicability Applies to storage facilities for new and spent fuel assemblies.
Objective To assure that both new and spent fuel assemblies will be stored in such a manner that an inadvertent criticality could not occur.
Specification 5.4.1 NEW FUEL STORAGE New fuel will nonrally be stored in the new fuel storage vault a.
or spent fuel pooit, The fuel assemblies are stored in racks __
l in parallel rows, 5 tving a nominal center to r: enter distance of 21-1/8 inches in both directions for the new fuel storage vault and the Spent Fuel Pool "A".
The fuel assemblies are stored in
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racks in pr.rallel rows, having a nominal center to center distance of 13-5/8 inches in both directiont for the Spent Fuel..
Pool "B".
This spacing is sufficient to maintain a K effective of less than.95 based on fuel assemblies with an enrichment of 4.3 weight percent U235 When fuel is being stored in the new fuel storage vault, twelve (12) storage locations (aligned in two rows cf six locations each; transverse row numbers four and eight) n.ust bc left vacant of fissile or moderating material to provide sufficient neutron leakage to satis'y the NRC maximum allowable reactivity value under the optimum low moderator density condition. When fuel is being moved in or over the Spent Fuel Storage Pool "A" and fuel is being stored in the pool, a boron concentration of at least 600 ppmb must be maintained to ensure meeting the NPC maximum allowable reactivity value under the postulated accident condition of a misplaced fuel assembly.
b.
New fuel may also be stored in the fuel transfer canal.
The fuel assemblies are stored in an 8 x 8 array storage rack having a nominal center to center distance of 21-1/8 inches.
When fuel is being moved in or over the fuel transfer canal, a boron concentration of at least 600 ppmb must be maintained to ensure that, under the postulated accident condition of a misplaced fuel assembly, the maximum reactivity will be less than the NRC maximum allowable reactivity. This applies only when fuel is being stored in the canal.
c.
New fuel may also be stored in shipping containers.
5-6 Amendmt:nt No.138
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5.4.2 SPENT. FUEL STORAGE i
Irradiated fuel assemblies will be stored, prior to offsite a.
i ship >nent, in the stainless. steel lined spent fuel pools, which are located is the fuel handling building, b.
Whenever there is fuel in the pool except for initial fuel loading, the spent fuel pool is filled with water borated to the concentration used in the reactor cavity and fuel transfer
- canal, Spent fuel may also be stored in storage racks in the fuel c.
transfer canal when the canal is at refueling level.
d.
The fuel assembly storage racks provided and the number of fuel elements each will store are listed by location below:
South End Spent Fuel Pool A Spent Fuel Pool B Dry New Fuel of Fuel North End of Fuel South End of Fuel Storage Area Transfer Handling Building Handling Building Fuel Handling Canal RB Building Fuel Assys 64
- 256 **
496 ***
66****
Cores 0.36 1.45 2.8 0.37 NOTES:
- Include space for accommodating a failed fuel detection container,
- Includes three spaces for acconsnodating failed fuel containers.
- Spent Fuel Pool B contains spent fuel storage racks with a reduced center-to-center spacing of 13 5/8 inches to increase the storage capacity of the pool.
- Includes twelve spaces which are required to be vacant of fissile or moderating material so that there is sufficient neutron leakage.
I All of the fuel assembly storage racks provided are designed to e.
Seismic Class 1 criteria to the accelerations indicated below:
Fuel Transfer Canal Fuel Handling Building Fuel Handling in Reactor Building Dry New Fuel Storage Area Building Spent And Spent Fuel Pool A Fuel Pool B Horiz.
0.76 g 0.38 g Vertical 0.?1 g 0.25 g 1
- The "B" pool fuel storage racks are designed using the floor response spectra of the Fuel Handling Building, f.
Fuel in the storage pool shall have a U-235 loading equal to or less than 57.8 grams of U-235 per axial centimeter of fuel assembly.
REFERENCES (1) FSAR Section 9.7 5-7 Amendment No. )6138
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