ML081200804
| ML081200804 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 04/18/2008 |
| From: | Dougherty T AmerGen Energy Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 5928-08-20072 | |
| Download: ML081200804 (17) | |
Text
AmerGen.S AmerGen Energy Company, LLC Telephone: 717-948-8ooo An Exelon Company Three Mile Island Unit i Route 441 South, P.O. Box 480 Middletown, PA 17057 April 18, 2008 5928-08-20072 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 THREE MILE ISLAND UNIT I (TMI UNIT 1)
OPERATING LICENSE NO. DPR-50 DOCKET NO. 50-289
SUBJECT:
BIENNIAL 10 CFR 50.59 AND COMMITMENT REVISION REPORT FOR 2006 AND 2007 Enclosed are the 2006-2007 Biennial 10 CFR 50.59 and Commitment Revision Report as required by 10 CFR 50.59 (d)(2) and SECY-00-0045 (NEI 99-04).
There are no regulatory commitments contained in this transmittal.
If you have any questions or require additional information, please contact Adam Miller, of Regulatory Assurance, at 717-948-8128.
Sincerely, Thomas J. Dougherty Plant Manager TJD/awm Enclosure cc: Administrator, Region I TMI-1 Senior Resident Irspector File 08027
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THREE MILE ISLAND UNIT 1 DOCKET NO. 50-298 BIENNIAL 10CFR 50.59 AND COMMITMENT REVISION REPORT TABLE OF CONTENTS 10CFR 50.59 Report 3 Modifications 4 Procedure Changes 9 Commitment Revision Report 13 2
AMERGEN ENERGY THREE MILE ISLAND UNIT I DOCKET NO. 50-289 BIENNIAL 10CFR 50.59 REPORT JANUARY 1, 2006 THROUGH DECEMBER 31, 2007 10CFR50.59 EVALUATION SUMMARIES 3
2006-2007 Biennial 10CFR 50.59 and Commitment Revision Report Modifications
Title:
RCS Zinc Injection Skid Installation Year Implemented: 2006 Evaluation Number: ECR 05-00631 Brief
Description:
This change installed a non-safety related vendor supplied zinc acetate chemical feed skid. The zinc injection program was implemented for the purpose of reducing the initiation of primary water stress corrosion cracking (PWSCC) of Reactor Coolant System (RCS) materials of construction and for reducing personnel exposure to irradiated corrosion product deposits. The zinc injection skid feeds the zinc acetate solution into the RCS via the normal makeup flow stream. At low concentrations, the zinc replaces radioactive cobalt, iron, and nickel in RCS component corrosion oxide layers. Over time, this reduces radiation fields and subsequently personnel radiation dose. At higher concentrations, the zinc is known to inhibit general corrosion and initiation of PWSCC. The impact of zinc addition will be addressed in the cycle specific Reload Safety Evaluation Report for each future cycle. These cycle specific analyses will ensure that all fuel rod design criteria are met with due consideration of zinc addition affects.
Summary of Evaluation:
The small flow and volume of zinc solution does not increase the frequency of operator response or the operator response time during boron dilution events.
The zinc injection has been evaluated such that there are no adverse effects on hydrogen generation or ECCS water pH and recirculation sump blockage. No fission product barriers are degraded. Radiation releases for planned and accident conditions have been evaluated with the conclusion that any increase due to zinc injection is bounded by existing analyses. Zinc is not the initiator of any accident, zinc does not introduce a new failure mode in fission product barriers, zinc will not change the manner in which the plant is operated, existing pipe break analyses bound any postulated breaks, zinc does not alter the response of the RCS or ECCS to transient conditions, and there is no effect on main Control Room. habitability.
In summary, the presently performed fuel and non-fuel analyses and the limitations, monitoring and/or testing that are imposed by future cycle specific Reload Safety Evaluation Reports demonstrate that zinc injection can be implemented without further evaluation. This change can be made without prior NRC approval.
4
2006-2007 Biennial I OCFR 50.59 and Commitment Revision Report
Title:
Removal of Decay Heat River Strainer (Temporary Change - ECR 06-00371, 06-00372)
Year Implemented: 2006 Evaluation Number: SE-000533-011 Brief
Description:
Perform a temporary configuration change to the decay heat river water system to remove the internals of the strainers DR-S-1A/B. The strainer motor assembly will be removed and disconnected electrically.
A significant increase in algae in the river water has caused rapid increase in the strainer's differential pressure due to the inability of the strainer blowdown to remove the material. This is a unique event that has not been previously experienced. The internals are being removed using a temporary change to eliminate this problem while the increase in algae in the river is occurring. This action is being taken to reduce the likelihood that the decay heat river water system will be challenged by this concern until a permanent change is developed to eliminate the concern.
The decay heat river water system is a safety related system that provides cooling to the decay heat closed cooling water system. Each train (A/B) of decay heat removal is 100% capable of performing the system's design function. Each train includes a decay heat river water system, a decay heat closed cooling water loop, and a decay heat removal system loop. The decay heat closed cooling water system provides cooling to the Emergency Core Cooling System (ECCS) pumps and cools the RCS in the reactor building sump recirculation emergency mode.
The decay heat closed cooling water system also cools the decay heat removal coolers during normal plant cooldown.
This temporary change has no overall adverse impact on plant operation. After the internals of the strainers are removed, the size of debris particles allowed to enter the system will increase. The total amount of debris that enters the system may also increase based on the environmental conditions of debris in the river. Based on size of the piping and components in the system relative to the size and amount of the debris that may enter the system after removing the internals of the strainers, the system will be capable of performing it design functions and meet all design basis performance criteria. No ECCS or other functions of the system are affected.
Monitoring of decay heat river system heat exchanger differential pressure and pump flow will be performed to verify that no significant degradation of the system occurs as a result of removing the strainer internals.
5
2006-200 7 Biennial I OCFR 50.59 and Commitment Revision Report Summary of Evaluation:
Based on the size of the piping and components in the system relative to the size and amount of the debris that may enter the system after removing the internals of the strainers, the system will be capable of performing its design functions and meet all design basis performance criteria. Implementation of this change does not create a common mode failure for the decay heat river water system trains.
This evaluation concludes that the change may be performed without prior NRC approval.
Title:
RPS Reactor Coolant Pump Power Monitor Trip Total Delay Time Year Implemented: 2007 Evaluation Number: ECR TM 06-00219 Brief
Description:
The delay time assumption used in TMI-1 safety analyses for the Reactor Protection System (RPS) reactor coolant pump power monitor (RCPPM) is revised from 620 msec to 930 msec. The RPS Design Basis Document (SDBD) and UFSAR are revised accordingly. No physical changes are made to the RCPPM hardware or settings. The delay time change is required to correct the safety analyses to be consistent with actual RPS hardware capabilities and surveillances. The current analyses of record are based on a shorter delay time of 620 msec.
Summary of Evaluation:
As discussed in UFSAR Section 14.1.2.6, the RCPPM is credited in the Loss of Coolant Flow accident, specifically the 4-0 pump coastdown. Reactor protection for the 4-0 pump coastdown event is provided by the power/pump monitors trip function of the RPS. The current TMI-1 analysis of record assumed a 620 msec total delay time for the RCPPM trip. A longer total delay time of 930 msec for the RCPPM trip would result in lower RCS flows in the core as the four reactor coolant pumps are coasting down while the reactor is still at power. This will result in lower minimum DNBR results for the 4-0 pump coastdown safety analysis, which is an adverse change.
Due to the adverse change in the total delay time assumed for the RCPPM trip in the 4-0 pump coastdown accident, Loss of Coolant Flow safety analysis, a 50.59 evaluation is required. The evaluation concluded that the change in the delay time assumption could be made without obtaining a License Amendment for the 6
2006-2007 Biennial I OCFR 50.59 and Commitment Revision Report following reasons: 1) Results from the revised 4-0 pump coastdown safety analysis using a 930 msec total delay time met the minimum DNBR acceptance criteria. The minimum DNBR design limit was not altered or exceeded. Since no fuels rods were predicted to depart from nucleate boiling, there were no changes to the consequences of the 4-0 pump coastdown accident. 2) The methodologies used to re-analyze the 4-0 pump coastdown accident are the same NRC-approved methodologies used in the current analysis of record. 3)
Determination of revised Maximum Allowable Peaking (MAPs) limits based on the new 4-0 pump coastdown results was performed in accordance with AREVA's NRC-approved topical report BAW-10179. 4) DNBR acceptance criteria for operating Cycle 16 are met with the revised MAPs using the current core operating limits.
Title:
Reanalysis of the Dropped Control Rod Accident Year Implemented: 2007 Evaluation Number: ECR No. 06-00550 Brief
Description:
In 2004, it was postulated by the fuel vendor (AREVA NP) that secondary thermal aging and embrittlement of the control rod drive mechanism (CRDM) leadscrew male coupling could lead to a complete failure of the bayonet during operation. If the bayonet were to fail, the control rod assembly (CRA) would fall into the core while the leadscrew remains in place. Since the position indicator is attached to the leadscrew the dropped rod would be undetected by the integrated control system/control rod drive control system (ICS/CRDCS) thus allowing the regulating rod group to be withdrawn. Failure of a CRDM leadscrew coupling was not postulated in the original plant safety analysis; therefore, a new analysis is required to evaluate the potential for a previously unanalyzed core power excursion.
This change adds a new malfunction to the control rod accident safety analysis for an event where the bayonet fails and the CRA falls into the core while the leadscrew remains in place. This safety analysis will be added to Section 14.1.2.7 of the UFSAR, which currently considers three types of CRA misalignment. This analysis does not change the conclusions previously provided in Section 14.1.2.7.
This activity has no impact on plant operations or design bases.
Summary of Evaluation:
For this modification, a 50.59 evaluation is required since the activity affects the UFSAR described design function. The proposed activity does not involve any physical changes or result in any adverse conditions. Since there are no 7
2006-2007 Biennial I OCFR 50.59 and Commitment Revision Report physical changes, there is no increase in the frequency of occurrence of accidents and no increase in the likelihood of occurrences of a malfunction. The results of the proposed activity demonstrate that the conclusions of UFSAR 14.1.2.7.e remain valid, i.e., the pressure criteria and departure from nucleate boiling ratio limits for the dropped rod analyses were both met. The proposed activity does not result in an increase in the consequences of a malfunction or accident and does not result in a design basis fission barrier being exceeded or altered. For this malfunction, AREVA NP analyses demonstrate that the ability of the reactor protection system trip function is not affected, no plant limits or margins are affected, and a License Amendment Request is not required. This activity creates changes to the UFSAR. This evaluation concludes that the change may be performed without prior NRC approval.
Title:
Design basis analyses changes for the response to Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors" Year Implemented: 2007 Evaluation Number: ECR No. 07-00743 Brief
Description:
This change implements the analyses to demonstrate compliance with the Generic Letter 2004-02. Two issues require a 50.59 evaluation.
The analysis associated with decay heat removal pump net positive suction head (NPSH), reduces the margin from 1.4 feet of water to 0.1 feet of water for the most limiting case. The reduction in margin is primarily the result of the conservative assumption that all accident generated debris is in place on the RB sump strainer at the start of recirculation mode Another analysis requires operator action to throttle flow from the Decay heat removal pumps to prevent exceeding the differential pressure limit of 16 feet of water on the reactor building sump strainer structure.
Summary of Evaluation:
The modification does not involve a change to any UFSAR described evaluation methodology or Technical Specifications (TS), and does not involve a test or experiment not described in the UFSAR. Decay Heat (DH) and Building Spray (BS) pumps' NPSH and flow requirements under the existing 50% debris blockage assumptions have been changed to meet the requirements of Generic 8
2006-2007 Biennial I0CFR 50.59 and Commitment Revision Report Letter (GL) 2004-02. The new and revised analysis being issued to support GL 2004-02 continues to consider the effects of debris clogging, minimum RB sump water level, and head loss for the maximum allowed flow rate plus instrument error. Throttling system flow to maintain NPSH margin is currently reflected in the design basis of the systems. Securing BS pumps after swap over to recirculation mode is being changed in the Emergency Operating Procedures.
The Building Spray operating times credited in the dose calculation C-1101-900-EOOO-087, "Post LOCA EAB, LPZ, TSC, and CR doses using AST and RG 1.183 Requirements," remains valid. UFSAR section 5.2.5.1, 5.2.5.2, 6.1.2, and 6.1.3 are affected by this activity.
The 50.59 evaluation for the reduction in NPSH margin and throttling LPI flow to limit differential pressure across the sump strainer determines that there is not a more than minimal increase in the likelihood of a malfunction of occurrence of a SSC important to safety previously evaluated in the UFSAR. The reduction in margin is primarily the result of the conservative assumption that all debris is present on the RB sump screen immediately when recirculation mode starts.
There is positive NPSH margin present for the most limiting case evaluated, which is one DHR pump and one building spray pump in service with the minimum reactor building cooling capability scenario. The addition of operator action to throttle flow from the DHR pumps to prevent differential pressure across the RB sump strainer from exceeding the design limit of 16 feet does not create a new failure mode and does not create a malfunction of a SSC important to safety with a different result than any previously evaluated because operator action to throttle flow from the DHR pumps currently exists in operating procedures and operators have indication available to verify that the correct action has been taken. This evaluation concludes that prior NRC approval is not required.
Procedure Changes
Title:
RPS Testing with Channel B in Manual Bypass Year Implemented: 2006 Evaluation Number: SE-000641-036 Brief
Description:
A reactor coolant system temperature element, RC4B-TE-2, has failed high resulting in a High RC temperature trip signal to "B" Reactor Protection System (RPS) Channel. "B" RPS Channel has been placed in Manual Bypass as a result of being inoperable due to RC4B-TE-2 failing. Per Technical Specification (TS)
Table 3.5-1, two channels are required to be operable with a minimum degree of redundancy of one. Currently there are two operable RPS channels with a 9
2006-200 7 Biennial I OCFR 50.59 and Commitment Revision Report degree of redundancy of one.
TMI-1 Technical Specification requires periodic surveillance testing of RPS Channels and Power Range Nuclear Instrument calibration. The surveillance test normally places the channel under test in Manual Bypass and substitutes test signals for the normal channel inputs. Channel B will be tested while it is in Manual Bypass. Channel A, C or D, RPS will be tested with the associated channel Tripped while Channel B remains in Manual Bypass. Tripping the channel under test while Channel B is in Bypass complies with Tech Spec Table 3.5-1, since two channels will be operable and one is tripped. Degree of redundancy of one is met with the channel under test in a tripped status. Testing the channel while it is in the Tripped status does not affect the validity of the test.
All the features are fully tested in accordance with Technical Specification Table 4.1-1.
This 50.59 evaluation supports changes to RPS related surveillance procedures as needed pending repair of RC4B-TE-2 in B RPS.
The design feature of the RPS is to allow for testing of a channel while in Manual Bypass, thus reducing the likelihood of an inadvertent reactor trip. Bypassing as described in UFSAR Chapter 7.1.2.3 h, allows bypassing one protection channel during on-line testing resulting in a 2 out of 3 coincidence. The adverse effect is that the RPS will be in a 1 out of 2 coincidence during testing of RPS Channels A, C, or D, thus increasing slightly the probability of reactor trip during testing.
Summary of Evaluation:
Testing the affected channel or calibrating the power range nuclear instrumentation with the channel tripped instead of in Manual Bypass does not affect the validity of the test, nor does it conflict with the UFSAR description. The system was designed to perform the testing while the channel is in bypass to reduce the risk of tripping the reactor. Testing with the channel tripped meets the TS Surveillance requirement and does not cause undue risk of a reactor trip.
These procedure changes do not result in more than a minimal increase to the frequency of occurrence of an accident previously evaluated in the UFSAR since the requirements of the LCO are met and the ability of the Reactor Protection System to respond to a valid condition are not compromised. The procedure change meets applicable NRC requirements as well as design standards. The procedures used provide steps to minimize the risk of tripping. The steps include
- 1) Verify no concurrent work activities on vital buses, 2) Verify no other testing or maintenance in progress or scheduled on RPS and 3) Verify RPS meets operability and degree of redundancy requirements of TS.
Reactor trip has been evaluated and the consequences of a reactor trip are not changed by these procedure changes although there is a slight increase in frequency of reactor trip and therefore events where reactor trip is the initiator.
10
2006-2007 Biennial JOCFR 50.59 and CommitinentRevision Report Operating with one channel tripped and one channel in bypass was found to be acceptable by the NRC staff during review of TS Amendment 189. The NRC concluded in its approval of TS Amendment 189 that there is no significant reduction in a margin of safety. Since the activity meets applicable NRC requirements and design standards there is no more than a minimal increase in frequency of occurrence of an accident from that previously evaluated in the UFSAR. This procedure change can be implemented without prior NRC approval.
Title:
Manual Auxiliary Fuel Handling Tool Operation Year Implemented: 2007 Evaluation Number: 1507-15 Brief
Description:
A new procedure, 1507-15, was developed to provide instructions for operation of a manual auxiliary fuel handling tool. The tool requires the use of the Fuel Handling Building (FHB) overhead crane to move irradiated fuel in the Spent Fuel Pool (SFP) at locations inaccessible to the fuel handling bridge and to allow certain Post Irradiation Exams (PIE) to be performed.
Summary of Evaluation:
The manual tool is specifically designed to handle Mark-B fuel, as is used at TMI-1.
The tool grapples identically to the tool used for new fuel receipt, and comparable to the current fuel bridge grapples. The grapple maintains a mechanical interlock, which prevents opening once loaded, the same as existing grapples. Additionally, the tool provides a hard stop to prevent raising an assembly above minimum shielding depth, similar to the existing bridge. To minimize the possibility of fuel handling damage, the procedure provides controls, consistent with procedure 1505-1, "Fuel and Control Component Shuffles," which ensure verification of grapple engagement, hoist position, load cell indications, hoist speed control, and rack position prior to executing the fuel movement steps. The Fuel Handling Supervisor maintains strict control of the fuel movements in accordance with the movesheets to prevent improper actions. Existing procedures and the UFSAR allow fuel movement with interlocks bypassed, with reliance on the administrative controls as discussed above. Additional controls are provided to prevent an uncontrolled malfunction of the crane from lifting an assembly out of the SFP. In addition, clamp on hard stops will be placed on both north/south rails and on the west side of the northern 3-ton crane trolley rail. The hard stops will prevent inadvertent crane motion from being able to pull a fuel assembly out of the water 11
2006-2007 Biennial 10CFR 50.59 and Commitment Revision Report after hitting a wall. These controls provide the basis for the determination that the fuel handling procedure changes do not constitute more than a minimal increase in the likelihood of a malfunction, which could damage irradiated fuel.
No increase in the frequency of occurrence or consequences of an accident will occur since the fuel handling accident in the SFP bounds any assembly drop accident during fuel handling. It would require the same amount of equipment failures and human errors to result in a fuel handling accident, whether using the fuel bridge or the manual tool, and the consequences would remain bounded by the previous UFSAR 14.2.2.1 analyses.
In summary, the new procedure may be implemented without prior NRC approval.
12
2006-2007 Biennial IOCFR 50.59 and Commitment Revision Report AMERGEN ENERGY THREE MILE ISLAND UNIT I DOCKET NO 50-289 BIENNIAL COMMITMENT REVISION REPORT JANUARY 1, 2006 THROUGH DECEMBER 31, 2007 13
2006-200 7 Biennial I OCFR 50.59 and Commitment Revision Report Letter Source: C311-89-2018, "Station Blackout Rule Response" Exelon Tracking No.: 5928-06-20500.001 Nature of Commitment: Testing requirements of the alternate AC power diesel generator (Station Blackout Diesel - SBO Diesel) that verify the SBO Diesel is capable of providing power to the safe shutdown bus within 10 minutes.
Summary of Justification:
This commitment has been deleted. These testing requirements are contained in plant procedures and have been incorporated into the UFSAR section 8.5.
Letter Source: 6710-96-2097, "Response to GL 89-13" Exelon Tracking No.: 6710-96-2097.001 Nature of Commitment: The commitment associated with this response includes the entire response letter. Page 5 of 8 of Attachment 1 of the letter contains information regarding the performance of maintenance and performance trending of the Control Building Chillers.
Summary of Justification:
The original commitment required the Control Building Chillers to be swapped every two weeks. The running frequency will be established on a frequency recommended by the equipment vendor. This level of detail is not required in the commitment.
Letter Source: NRC Inspection Report 1980-22 Exelon Tracking No.: 1980T0033 Nature of Commitment: Required the development and implementation of approved training and replacement training programs as implemented by procedures.
Summary of Justification:
This commitment has been deleted. The procedures referenced were updated prior to TMI-1 Restart. This satisfied the commitment. The training procedures/processes were revised based on the new INPO Accreditation 14
2006-2007 Biennial I OCFR 50.59 and Commitment Revision Report requirements that were instituted in 1985. TMI-1 meets the INPO ACAD requirements.
Letter Source: NRC letter 5211-83-3232: SER ON INSTRUCTOR INDOCTRINATION AND QUALIFICATION TRAINING Exelon Tracking No.: 1983T0152 Nature of Commitment: Required the development and implementation of approved training and replacement training programs as implemented by procedures.
Summary of Justification:
This commitment has been deleted. The procedures referenced were updated prior to TMI-1 Restart. This satisfied the commitment. The training procedures/processes were revised based on the new INPO Accreditation requirements that were instituted in 1985. TMI-1 meets the INPO ACAD requirements.
Letter Source: NRC Inspection Report 1980-19 Exelon Tracking No.: 1980T0025 Nature of Commitment: Committed to the development of a training department administrative manual.
Summary of Justification:
This commitment has been deleted. The training department administrative manual was established prior to TMI-1 Restart. This satisfied the commitment.
The training department currently complies with the INPO ACAD 02-001 and ACAD 02-002 requirements for accreditation of station training programs that were put in place when INPO began the accreditation process in 1985.
Letter Source: LER 91-004 Exelon Tracking No.: 1991T0046 Nature of Commitment: Committed to revise operator lesson plans by including the LER event: "Movement of an irradiated fuel 15
2006-2007 Biennial I0CFR 50.59 and Commitment Revision Report assembly without having containment integrity due to operator errors and procedural weaknesses."
Summary of Justification:
This commitment has been deleted. TMI-1 Technical Specification 3.8.6 now states: "During the handling of irradiated fuel in the Reactor Building at least one door in each of the personnel and emergency air locks shall be capable of being closed.
- Administrative controls shall ensure that the Reactor Building Purge Exhaust System is in service, appropriate personnel are aware that air lock doors and/or other penetrations are open, a specific individual(s) is designated and available to close the air lock doors and other penetrations as part of a required evacuation of containment." As a result, movement of irradiated fuel is now allowed without having containment integrity, which was the original subject of the LER, and therefore this commitment is no longer applicable.
Letter Source: IEB 74-8: DEFICIENCY IN ITE MOLDED CASE CIRCUIT BREAKERS, TYPE HE-3 Exelon Tracking No.: 1974T0015 Nature of Commitment: Committed to additional testing of HE-3 type ITE molded case circuit breakers (MCCBs) are conducted at least once every refueling period.
Summary of Justification:
This commitment has been deleted. TMI trip tests the ITE Type HE-3 MCCBs in accordance with corporate "best practices" PCM template for Motor Control Centers/MCCBs on a frequency of 6 years for "critical" components and 10 years for "non-critical components. All trip testing is performed in accordance with industry methodology (NEMA Standard AB-1-2000 as implemented by the applicable site procedures. Testing of ITE HE-3 Molded Case Circuit Breakers is performed in accordance with Exelon Nuclear Procedures E62.1 entitled "Molded Circuit Breaker Testing-Thermal Magnetic Trip" and E62.2 entitled "Molded Case Circuit Breaker Testing-Instantaneous Trip."
Letter Source: 6710-96-2097, "Response to GL 89-13" Exelon Tracking No.: 6710-96-2097.001 16
2006-200 7 Biennial I OCFR 50.59 and Commitment Revision Report Nature of Commitment: The commitment associated with this response includes the entire response letter. Page 4 of 8 of Attachment 1 of the letter contains information regarding the inspection of the water side of the Reactor Building Air Coolers.
Summary of Justification:
There are three Reactor Building Air Coolers. The original commitment required that inspection of one cooler's water side would occur each outage. A review of water side cooler inspections from 1994 to 2005 found the tubes in excellent condition, i.e. tube ID had no blockage, no debris, and no fouling. The tubes are expected to be in this condition because they are normally in wet lay-up with treated demineralized water. GL 89-13 is concerned with fouling in service water coolers. Since these cooling coils are in a controlled chemistry wet lay-up condition most of the time, and the coolers are tested to verify accident design flow rate, there is no reason to open and inspect a sample of tube inner diameters (IDs). Inspection of the tube Ids have been performed three times for each cooler and all inspections showed excellent tube ID conditions. Therefore this commitment is changed to delete inspection of one cooler's water side each refueling inspection.
Letter Source: 6710-97-2023, "Response to NRC pursuant to 10 CFR 50.54(F) concerning System Design Bases Documentation" Exelon Tracking No.: 96082.01 Nature of Commitment: This commitment originally referenced GPUN Engineering procedure EP-045, "Review and Control of System Design Basis Documents." Subsequently this procedure was superceded by Exelon procedure CC-AA-207, "Control of Design Basis/Baseline Documents and Design Basis Databases."
Summary of Justification:
This commitment is being changed to reference the corporate procedure, CC-AA-207, in place of the originally referenced site procedure, EP-045. The scope of guidance in the corporate procedure is similar to the previously referenced site procedure and therefore there is no change of intent in this commitment change.
End of Commitment Revision Report 17