ML20057A830
| ML20057A830 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 09/10/1993 |
| 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 | |
| ML20057A831 | List: |
| References | |
| DPR-50-A-178 NUDOCS 9309150393 | |
| Download: ML20057A830 (9) | |
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- ( (f; j 9 UNITED STATES (sfg j
NUCLEAR REGULATORY COMMISSION s'
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,e WASHINGTON. D C. 20555-0001 y
METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT COMPANV l
PENNSYLVANIA ELECTRIC COMPANY i
GPU NUCLEAR CORPORATION
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DOCKET N0. 50-289 THREE MILE ISLAND NUCLEAR STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 178 i
t License No. OPR-50 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by GPU Nuclear Corporation, et al.
(the licensee), dated June 7, 1993, complies with the standards and
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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
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this amendment can be conducted without endangering the health and i
i 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.
i 9309150393 930910 ADOCK0500{9 i
PDR P
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Accordingly, the license is amended by. changes to the Technical Specifications as indicated in the attachment to this license i
amendment, and paragraph 2.c.(2) of Facility Operating License No.
j DPR-50 is hereby amended to read as follows:
i (2) Technical Specifications The Technical Specifications contained in Appendix A, as l
revised through Amendment No.178, are hereby incorporated in i
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, to be h
implemented within 60 days of issuance.
l FOR T NUCLEAR REGULATORY COMMISSION i
n 9
John F. Stolz, Di ector Project Directorate I-4 i
Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation j
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Attachment:
l Changes to the Technical Specifications I
Date of Issuance: Septenber 10, 1993 l
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A_TTACHMENT TO LICENSE AMENDMENT N0.178 FACILITY OPERATING LICENSE NO. DPR-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 3-21 3-21 3-23 3-23 3-24 3-24 3-45 3-45 5- -4 5-4 6-19 6-19 l
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3.3 EMERGENCY CORE COOLING. REACTOR BUILDING EMERGENCY COOLING AND REACTOR BUILDING SPRAY SYSTEMS applicability Applies to the operating status of the 2mergency core cooling, reactor building emergency cooling, and reactor building spray systems.
Ob.iective To define the conditions necessary to assure immediate availability of the emergency core cooling, reactor building emergency cooling and reactor building spray systems.
Soecification 3.3.1 The reactor shall not be made critical unless the following conditions are met:
3.3.1.1 In.iection Systems a.
The borated water storage tank shall contain a minimum of 350,000 gallons of water having a minimum concentration of 2,500 ppm boron at a temperature not less than 40'F.
Specification 3.0.1 applies.
b.
Two makeup pumps are operable in the engineered safeguards mode powered from independent essential buses.
Specification 3.0.1 applies.
c.
Two decay heat removal pumps are operable.
Specification 3.0.1 applies.
d.
Two decay heat removal coolers and their cooling water supplies are operable. (See Specification 3.3.1.4)
Specification 3.0.1 applies.
e.
Two BWST level instrument channels are operable.
f.
The two reactor building sump isolation valves (DHV6A/B) shall be either manually or remote-manually operable.
Soecific.ation 3.0.1 applies.
3.3.1.2 Core Floodina System Two core flooding tanks each containing 1040 i 30 f t' of a.
borated water at 600 25 psig shall be available.
Specification 3.0.1 applies.
b.
Core flooding tank boron concentration shall not be less than 2,270 ppm boron.
The electrically operated discharge valves from the core flood c.
tank will be assured open by administrative control and position indication lamps on the engineered safeguards status j
panel. Respective breakers for these valves shall be open and i
conspicuously marked.
Specification 3.0.1 applies.
d.
One core flood tank pressure instrumentation channel and one core flood tank level instrumentation channel per tank shall be operable.
i Amendment No. ?!, 93, 178 3-21
3.3.3 Exceptions to 3.3.2 shall be as follows:
1 a.
Both core flood tanks shall be operable at all times.
b.
Both the motor operated valves associated with the core flood tanks shall be fully open at all times.
c.
One reactor building cooling fan and associated cooling unit shall be permitted to be out-of-service for seven days.
3.3.4 Prior to initiating maintenance on any of the components, the duplicate (redundant) component shall be verif'ed to be operable.
Bases The requirements of Specification 3.3.1 assure that, before the reactor can be made critical, adequate engineered safety features are operable.
Two engineered safeguards makeup pumps, two decay heat removal pumps and two decay heat removal coolers (along with their respective cooling water systems components) are specified.
However, only one of each is necessary to supply emergency coolant to the reactor in the event of a loss-of-coolant accident.
Both core flooding tanks are required because a single core flooding tank has insufficient inventory to reflood the core for hot and cold line breaks (Reference 1).
4 The operability of the borated water storage tank (BWST) as part of the ECCS i
ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA (Reference 2). The limits on BWST i
minimum volume and boron concentration ensure that 1) sufficient water is available within containment to permit recirculation cooling flow to the core, and 2) the reactor will remain at least one percent subcritical following a loss-of-Coolant Accident (LOCA).
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The contained water volume limit of 350,000 gallons includes an allowance for water not usable because of tank discharge location and sump recirculation switchover setpoint. The limits on contained water volume, NaOH concentration and boron concentration ensure a pH value of between 8.0 i
and 11.0 of the solution sprayed within containment after a design basis accident.
The minimum pH of 8.0 assures that iodine will remain in solution while the maximum pH of 11.0 minimizes the potential for caustic damage to mechanical systems and components.
Redundant heaters maintain the borated water supply at a temperature greater than 40'F.
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3-23 Amendment No. Jyp, J57, J5), 178
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The post-accident reactor building emergency cooling may oe accomplished by
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three emergency cooling units, by two spray systems, or by a combination of i
one emergency cooling unit and one spray system. The specified requirements i
J assure that the required post-accident components are available.
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The iodine removal function of the reactor building spray system requires J
one spray pump and sodium hydroxide tank contents.
j The spray system utilities common suction lines with the decay heat removal I
system.
If a single train of equipment is removed from either system, the other train must be assured to be operable in each system.
When the reactor is critical, maintenance is allowed per Specification 3.3.2 and 3.3.3 provided requirements in Specification 3.3.4 are met which assure i
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operability of the duplicate components. The specified maintenance times i
are a maximum. Operability of the specified components shall be based on the satisfactory completion of surveillance and inservice testing and 1
inspection required by Technical Specification 4.2 and 4.5.
The allowable maintenance period of up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> may be utilized if the operability of equipment redundant to that removed from service is verified based on the results of surveillance and inservice testing and inspection required by Technical Specification 4.2 and 4.5.
In the event that the need for emergency core cooling should occur, operation of one makeup pump, one decay heat removal pump, and both core i
flood tanks will protect the core.
In the event of a reactor coolant system i
rupture their operation will limit the peak clad temperature to less than i
i 2,200*F and the metal-water reaction to that representing less than I l
i percent of the clad.
l Two nuclear service river water pumps and two nuclear service closed cycle cooling pumps are required for normal operation. The normal operating l
requirements are greater than the emergency requirements following a J
l os s-o f-cool ant.
i REFERENCES (1) UFSAR, Section 6.1
" Emergency Core Cooling System" i
i (2) UFSAR, Section 14.2.2.3 "Large Break LOCA" 1
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Amendment No. $7, 149, J57, Jgy, 178
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i 3.8.9 The reactor building purge system, including the radiation monitors which initiate purge isolation, shall be tested and verified to be operable no more than one week prior to refueling operations.
3.8.10 Irradiated fuel shall not be removed from the reactor until the unit has been subcritical for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Bases Detailed written procedures will be available for use by refueling personnel.
These procedures, the above specifications, and the design of the fuel handling equipment as described in Section 9.7 of the UFSAR incorporating built-in interlocks and safety features, provide assurance that no incident could occur during the refueling operations that would result in a hazard to public health and safety.
If no change is being made in core geometry, one flux monitor is sufficient. This permits maintenance on the instrcmentation.
Continuous monitoring of radiation levels and neutron flux provides imediate indication of an unsafe condition.
The decay heat removal pump is used to maintain a uniform boron concentration.
The shutdown margin indicated in Specification 3.8.4 will keep the core subcritical, even with all control rods withdrawn from the core (Reference 1).
The boron concentration will be sufficient to maintain the core k,,, 1 0.99 if all the control rods were removed from the core, however only a few control rods will be removed at any one time during fuel shuffling and replacement.
The k,, with all rods in the core and with refueling boron concentration is approximately 0.9.
Specification 3.8.5 allows the control room operator to inform the reactor building personnel of any impending unsafe condition detected from the main control board indicators during fuel movement.
The specification requiring testing Reactor Building purge termination is to verify that these components will function as required should a fuel handling accident occur which resulted in the release of significant fission products.
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Specification 3.8.10 is required as the safety analysis for the fuel handling accident was based on the assumption that the reactor had been shutdown for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (Reference 2).
REFERENCES (1) UFSAR, Section 14.2.2.1
" Fuel Handling Accident" i
l (2) UFSAR, Section 14.2.2.l(2)
" FHA Inside Containment" 3-45 Amendment No. J)7,178
5.3 REACTOR Acolicability Applies to the design features of the reactor core and reactor coolant system.
Ob.iective To define the significant design features of the reactor core and reactor coolant system.
Soecification 5.3.1 REACTOR CORE 5.3.1.1 The reactor core is composed of slightly enriched uranium dioxide pellets contained in fuel rods.
A fuel assembly normally contains 208 fuel rods arranged in a 15 by 15 lattice.
The details of the fuel assembly design are described in TMI-1 UFSAR Chapter 3.
5.3.1.2 The reactor core shall approximate a right circular cylinder with an equivalent diameter of 128.9 inches. The active fuel height is defined in THI-1 UFSAR Chapter 3.
5.3.1.3 The core average and individual batch enrichments for the present cycle are described in TMI-l UFSAR Chapter 3.
5.3.1.4 The control rod assemblies (CRA) and axial power shaping rod assemblies (APSRA) are distributed in the reactor core as shown in THI-l FSAR Chapter 3.
The CRA and APSRA design data are also described in the UFSAR.
5.3.1.5 The THI-l core may contain burnable poison rod assemblies (BPRA) and gadolinia-urania integral burnable poison fuel pellets as described in THI-l UFSAR Chapter 3.
5.3.1.6 Reload fuel assemblies and rods shall conform to design and evaluation data described in the UFSAR.
Enrichment shall not exceed a nominal 5.0 weight percent of U,.
i 5.3.2 REACTOR COOLANT SYSTEM 5.3.2.1 The reactor coolant system shall be designed and constructed in accordance with code requirements.
(Refer to UFSAR Chapter 4 for details of design and operation.)
5.3.2.2 The reactor coolant system and any connected auxiliary systems exposed to the reactor coolant conditions of temperature and pressure, shall be designed for a pressure of 2,500 psig and a 0
temperature of 650 F.
The pressurizer and pressurizer surge line 0
shall be designed for a temperature of 670 F.
5-4 Amandment No. Jgp, JJZ, J59, J57, J79,178
6.9.5 CORE OPERATING LIMITS REPORT i
6.9.5.1 The core operating limits addressed by the individual Technical l
Specifications shall be established and documented in the CORE OPERATING LIMITS REPORT prior to each reload cycle or prior to any remaining part of a reload cycle.
6.9.5.2 The analytical methods used to determine the core operating limits addressed by the individual Technical Specifications shall be those previously reviewed and approved by the NRC for use at i
THI-1, specifically:
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(1) BAW-10179 r -A, " Safety and Methodology for Acceptable Cycle r
Reload Analyses." The current revision level shall be l
specified in the COLR.
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I 6.9.5.3 The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core j
thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient / accident analysis limits) of the safety analysis are met.
6.9.5.4 The CORE OPERATING LIMITS REPORT, including any mid-cycle I
revisions or supplements thereto, shall be provided upon issuance for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
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6-19 Amendment No. 72, 77, JZS, JJ7, JJJ, JJS, J57, J5$, J77,178
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