ML20212R276
| ML20212R276 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/20/1987 |
| From: | Hukill H GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| 5211-87-2076, NUDOCS 8704270046 | |
| Download: ML20212R276 (19) | |
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,4 GPU Nuclear Corporation NggIgf Post Office Box 480 Route 441 South Middletown, Pennsylvania 17057-0191 717 944 7621 TELEX 84 2386 Writer's Direct Dial Number:
April 20 1987 5211-87 $076 US Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
Dear Sir:
Three Mile Island Nuclear Station, Unit 1 (TMI-1)
Operating License No. DPR-50 Docket No. 50-289 10 CFR 50.59 Report for 1986 In accordance with the requirements of 10 CFR 50.59, enclosed are summaries of the changes to THI-1 systems and procedures as described in the Safety Analysis Report.
Sincerely,
./
. D.
u ill Vice President & Director, TMI-1 HDH/SM0/spb:0844A cc:
J. Allan, USNRC R. Conte, USNRC i
b 8704270046 870420 l(O PDR ADOCK 05000289
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PDR p
GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation
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Modification:
125/250V.DC Station Battery "A" Replacement (B/A-412411)
Description of Modification:
This modification upgraded the 125/250V DC system by replacing the "A" station battery. The existing battery was nearing the end of its expected life due to t
capacity decline as a result of aging. The' replacement battery has a greater load capacity (ampere-hours) to accommodate 'present load, load growth, and
- capacity decline due. to aging. The load capacity has been increased from 1200 i
to 1500 ampere-hours.
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Safety Evaluation Summary:
The new battery has a greater ampere-hour rating which will decrease the probability of failure due to age or overloading. The DC Distribution System and its connections to the batteries have not been modified and are of sufficient size to handle the additional battery rating. The longer recharge time necessary due to the larger battery has been determined to be acceptable.
Replacement of the battery did not increase the possibility of occurrence or consequences of an accident or malfunction, or reduce a safety margin.
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typtification:
Replacement of Snubbers'with' Rigid Struts
- Description of Modification:
This modification replaced the _ following list of. snubbers with rigid struts to provide the necessary seismic restraint to TMI piping:
Snubber Mark No.:
DCH-58, 59 EF-74, 75, 76, 85, 89 IPE-2 MUE-20
- NSE-2, 3, 7, 8, 9,10,12,13,14,15,16,17,18, 23, 24, 27, 29, 33 4
RWE-8, 9,10,11,12,13,14 RW-72, 73, 75 3
Safety Evaluation Summary:-
~ The system performance of the piping is not affected by replacement of snubbers with rigid struts since as seismic restraints, they perform =
identically as snubbers.
Replacement of the snubbers with rigid struts does not alter thermal analysis, because only snubbers on systems with minimal thermal growth were replaced.
No plant or system operational changes have been made. and original design acceptance criteria safety margins are still satisfied, therefore, no unreviewed safety question exists.
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Modification:
MS-V-22A/B' Replacement (CMR #0645M)
Description of Modification:
Main Steam Safety Valves MS-V22A/228 (Lonergan Model V-618) that had set.
pressures of 200 psi and 220 psi have been replaced with new Lonergan Valve Model No. S30P that have higher set pressures of 260 psi.and 280 psi correspondingly. The higher relief pressures wil1~ increase reliability of
' EF-P1 operation without lifting MS-V22A/228. This increases the operating l
margin between the turbine operating pressure and the safety valve relief setting.
Safety Evaluation -Summary:
The activity of installing new MS-V-22A/B safety relief valves with increased setpoints of 260 and 280 psig by using spacer rings is technically acceptable and does not constitute an unreviewed safety concern as defined in 10 CFR 50.59.
This conclusion is based on:
I a.
It does not alter EFW system actuation, operation, control or design conditions or features.
f b.
It has no detrimental affect on the main steam syst'em.
1 c.
It enhances the EF-P-1 turbine operation by increasing the reliability of the steam supply.
d.
Over-pressure protection is maintained within an acceptable range for all associated piping and components.
The change has been evaluated and shown that it meets the required e.
piping, seismic and applicable design criteria, f.
Does not require operator action to perform their intended function or a change to existing operating or emergency proc'edures, except to reflect
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Modification:
RCDT " Pop-Off" Plug (CMR #0571M).
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Description of Modification:
. A " pop-off" plug was installed to reduce back leakage from the Reactor Building to the Reactor Coolant Drain. Tank if the Reactor Building pressurizes.
Safety Evaluation Summary:
i The " pop-off" plug has been incorporated as a permanent modification to mitigate the consequences of any Reactor Building atmosphere pressurization that may occur in the future.
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. Installation of the " pop-off" plug will not have an adverse effect on the operation of the RCDT vent line or any other ITS system.
The function of the tank vent line is unaffected.
This change does not result in an increase of the probability or possibility of an accident nor does it reduce the margin of safety as defined in the SAR and Tech. Specs.
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j Modification:
Removal of Instrument Air to Valves EF-V-8A/B/C (B/A 123170)
Description of Modification:
This modification removed the air supply tubing from the Instrument, Back-Up, and 2 Hour Back-Up Air Supplies to the diaphram operators.
This precludes the possibility of the EF.-V-8A/B/C valves from being closed, thus retaining the minimum recirculation flow paths for protection of the emergency feedwater pumps EF-P-1/2A/2B during and after a seismic event.
The EF-V-8A/B/C valves are in a failed open position and mechanical stem blocks are installed.
Safety Evaluation Summary:
The removal does not affect the Emergency Feedwater System from performing it's intended function. The EF-V-8A/B/C valves being in their open position, has been previously evaluated to assure that there still ~ remains enough water flow to the OTSGs so as to satisfy the design basis of the Emergency Feedwater System.
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Hodification:
Removal of Back-Up Air to EF-V-8A (CMR #0515M)
Description of Modification:
Instrument air tubing between IA-V-1125 and the T-fitting connection to EF-V-8A was capped at both ends.
It had previously been taken out of service by an Emergency Feedwater Upgrade modification.
j Safety Evaluation Summary:
Removal of back-up air to EF-V-8A does not constitute an unreviewed safety question as defined in 10 CFR 50.59. This modification was evaluated as part of an earlier modification to block open EF-V8A/B/C and disconnect' instrument air from them.
Removal and capping of the back-up air lines to EF-V-8A is not a safety concern and will not affect the ability of the Emergency Feedwater System to perform its intended function.
Modification:
Condenser Off-Gas Sampling System (B/A 412452)
Description of Modification:
The main and auxiliary vacuum pumps remove non-condensable gases from the main condenser to maintain a sub-atmospheric pressure in the main condenser. The removed non-condensables are discharged to a common header that discharges to.
the atmosphere. This modification provides the capability' to continuously sample the condenser vacuum pump exhaust (off-gas) for potential radioactive effluent. The sample is obtained from.the vacuum pump discharge header where it is directed to a sample chamber that collects both particulates and fodines from the sample flow stream via.a HEPA filter and charcoal canister. The-sample flow stream is returned to the main condenser vacuum pump suction line. At prescribed intervals, the sample chamber is removed for laboratory analysis.
Safety Evaluation Summary:
1 The Condenser Off-Gas Sampling System does not affect the system performance of the condenser or condenser vacuum pumps.
The modification was designed and fabricated to the design requirements at TMI-1.
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As a result, the probability of occurrence or the consequence of an accident or malfunction of equipment previously evaluated in the Safety Analysis Report -
has not been increased.
It is, therefore, concluded that the subject modification does not involve an unreviewed safety question per the criteria of 10 CFR 50.59.
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Modification:
Reactor Purge, Auxiliary / Fuel Handling Building Exhaust Makeup Air (B/A 412383)
Description of Modification:
This modification installed a plenum wall in the TMI-l Air Intake Tunnel, a 48" round duct running from upstream of the Air Intake Filters to the-plenum wall, and a 48" round duct running from the Air Intake Tunnel to the Reactor Building Purge and Auxiliary / Fuel Handling Building exhaust makeup air dampers located in the Ventilating Equipment Room, elevation 305' in the Auxiliary Building.
In addition, the makeup air dampers were replaced with new dampers and controllers sized for the revised system flowrates.
The purpose of this modification is to provide direct ducting of makeup air from upstream of the Air Intake Filters to the makeup air dampers thereby maintaining minimum flow requirements for the Reactor Building Purge and Auxiliary / Fuel Handling Building exhaust fans, eliminating negative pressure problems in the Ventilating Equipment Room, and preventing the possibility of drawing contaminants directly into the Auxiliary Building exhaust, bypassing the exhaust filters.
Safety Evaluation Summary:
The engineering and installation of this modification were classified Important to Safety because it required installation of concrete anchors into a Class I structure (the Auxiliary Building) and due to the requirement that the duct supports and plenum wall meet the seismic Class II criteria for missile hazards. The duct work itself is classified Not Important to Safety.
The probability of occurrence or the consequences of an accident or malfunction of equipment previously evaluated in the Safety Analysis Report has not been increased.
This modification does not involve any unreviewed safety question per the criteria of 10 CFR 50.59.
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Modification:
Control Room Habitability Upgrade Anmonium Rydroxide Tank Dike Expansion (B/A 412168)
Description of Modification:
This project upgraded the Dike around the TMI-1 Ammonium Hydroxide Tank so that the entira contents of the tank, if full, would be confined within the i
dike if a spill were to occur. The confinement of a spill improves Control Room Habitability so that the evaporative surface area of the gas and resulting concentration of anmonia in the control room are within acceptable limits.
Safety Evaluation Summary:
For modifications to the dike described above,. it is concluded that:
1.
The probability of consequences of accidents previously evaluated have not increased since this modification is intended to preclude Control Room personnel loss of function due to ammonium hydroxide tank rupture.
2.
This modification will not increase the probability of occurrence or consequences of a malfunction of equipment ITS because only relocation of the eyewash station and fuel oil lines will take place, which will not affect these systems' performance.
l 3.
No accidents other than those previously considered will be introduced, as no systems are modified to the point where system performance is affected.
4 This modification does not decrease the margin of safety as defined in the basis of plant technical specifications because it is adding a safety provision that was previously non-existent and it does not affect system performance.
Based on the above conclusions, the modifications to the Ammonium Hydroxide Dike described in this report do not involve an unreviewed safety '
consideration with regard to the criteria of 10 CFR 50.59.
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Modification:
RC-V-1,- RC-V-3 and MU-V-1 A
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i Packing Leak-Off Line Elimination (CMR #0462)
Description of Modification:
Steam leak-off connections for RC-V-1, 3 and MU-V-1 A were cut and capped.
Safety Evaluation Summary:
This modification made the affected valves more reliable and more maintainable by making a packing configuration standard which is more effectively compressed. Any leakage from the modified packing chambers will now go to atmosphere rather than to the RC drain tank but the leakage is far less likely and can be detected and eliminated in a timely manner.
The elimination of packing leak-off is therefore determined to be an improvement in nuclear safety and is consistent with the NSR/ITS classification of the valves and piping systems which are affected.
Modification:
Fuel' Handling Building Crane - Revised Heavy Load Travel Path (B/A 412481)
Descript;[a of Modification:
This modification repositioned the Fuel Handling Building (FHB) crane limit switches that govern the travel path limits of the crane during heavy load
(>15 tons) operation. These travel limits have been reset to meet the GPUN commitnent to the NRC in response to NUREG-0612 " Handling of Heavy Loads."
The only new components added per this modification are hardware items for mounting the existing limit switches in their new positions.
Safety Evaluation Summary:
The arrangement of the FHB crane travel limit switches was changed to further restrict crane travel with heavy loads.
The revised travel path. limits reduce the potential for impacting irradiated material in the spent fuel pool. The function of the relocated limit switches remains as originally designed.
Thus, this modification does not reduce the margin of safety of the TMI-1 plant as described in the FSAR.
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Modification:
Permanent Flow Indicator for Nuclear Services Flow Element (CMR #0306M).
Description of Modification:
i c A permanent flow indicator was added Eto flow element NS-FX-232. The change involved running instrument tubing from the existing flow element to a new, locally mounted, flow' indicating instrument.
Safety Evaluation Summary:
-Installing the permanent flow indicator does not. increase the probability of occurrence or the consequences of an accident or malfunction of equipment since it allows continuous monitoring of Nuclear Services flow to the heat exchangers. The flow element already existed to provide a differential pressure to measure the flow periodically for test purposes. Adding the permanent gauge makes it convenient for test personnel to read flow. The margin of safety is not reduced since the ability to measure Nuclear Services i
flow remains.
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Modification:
CA7-FI Relocation (CMR #0623M)
Description of Modification:
Flow indicator CA7-FI for the Chemical Addition System was relocated in order to provide greater. visibility to Operations and Plant Chemistry personnel.
Changes involved removal from existing location downstream of CA-RV5 and reinsta11ation just upstream of CA-RYS.
Safety Evaluation Summary:
This change relocates the existing CA-7-FI rotameter to an area which provides greater visibility to plant personnel.
The flowmeter was reinstalled in accordance with criteria which equal or exceed the original criteria. No licensing or operational changes are required as a result of this change.
The change creates no safety concerns.
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Modification:
Long Term Human Factors Modifications of Control Room (B/A 412381)
Description of Modification:
This modification consisted of incorporating human factors standards to Panel PL in the control room. The 120 volt valve indicator lamps were redesigned for the standard 24 volt microswitch indicator lamps. The mimic
. flow path for the Hydrogen Analyzer and for the Hydrogen Recombiners were incorporated into the panel. This modification was installed on an insert which was then installed in Panel PL. This modification was minor in nature and resulted in the same design configuration as existed prior to the change.
Safety Evaluation Summary:
I The equipment that was replaced is fully compatible with equipment in the plant. The functions and basic configuration remain as originally designed.
Thus, this modification does not reduce the margin of safety of the TMI-1 plant as described in the FSAR.
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Modification:
0TSG "A" Inspection Ports (B/A 128079)
Description of Modification:
This modification installed two 4-3/4" inspection openings in the "A" Once Through Steam Generator (OTSG "A").
An opening was installed between the 3rd and 4th tube support plates and another was installed between the 5th and 6th tube support plates. The new openings facilitate inspection and sampling of the potential debris buildup area.
Safety Evaluation Summary:
The proposed size and locations of the new inspection holes sat.isfy the applicable ASME Code requirements and are bounded by the existing stress analysis. This modification does not affect the steam generator integrity-and performance and does not involve an unreviewed safety question.
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Procedure:
1000-ADM-1291.01 Procedure for Nuclear Safety and Environmental Impact Review and Approval of Documents Description of Change:
Previous versions of this procedure specified that change's' to procedures or plant equipment classified as Important to Safety required a written Safety Evaluation and Independent Safety Review. This procedure was revised to delete the Important to Safety classification criterion and to provide a checklist of screening criteria to determine whether a written Safety Evaluation and consequent Independent Safety Review were required.
Safety Evaluation Summary:
This procedure revision deletes Important to Safety (ITS) as a determinant of whether this procedure applies to a document.
Other determinations which focus on potential impacts on nuclear safety as well as the regulatory required 10 CFR 50.59 considerations are judged to be more important in assessing the need for a detailed written safety evaluation. The procedure continues to ensure adherence to Technical Specification requirements.
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1101 Plant Limits and Precautions Description of Change:
The maximum allowed temperature of the contents of the Borated Water Storage Tank was changed from 90*F to 120*F.
j Safety Evaluation Summary:
The LOCA consequences of operating the TMI-l unit at 100% power with a BWST temperature of 120*F have been evaluated.
The present generic large break LOCA analyses will bound the minimal impact of the 30*F elevation in BWST temperature. As for generic small break LOCA evaluations, the higher BWST temperature will result in a decreased vessel inventory. However, plant specific calculations demonstrate that the power level difference between the generic plant and TMI-l is more than adequate to offset the effects of the elevated BWST temperature. Mass and energy releases to.the containment would increase by 0.35% due to _the higher BWST temperature; however, the resultant increase in building pressure would be less than 0.5 psi.
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Procedure:
1101-2 -- Plant Setpoints Description of Change:
The Sodium Hydroxide Tank level high alarm setpoint was changed from 46.3 feet to 47.2 feet.
Safety Evaluatic Summary:
The new setpoir,t allows a margin of 1.5 feet to account for, possible instrument leap errors and to stop the tank filling process. Containment spray requirements specified by Technical Specifications are governed by tank concentration and level difference with the BWST.
This specification uses separate differential pressure instrumentation.
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