ML051890044
| ML051890044 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 06/29/2005 |
| From: | Price J Dominion Nuclear Connecticut |
| To: | Document Control Desk, NRC/FSME |
| References | |
| 05-399 | |
| Download: ML051890044 (22) | |
Text
Dominion Nuclear Connecticut, Inc.
Millstone Power Station vO Dominion Rope Ferry Road Watcrford, CT 06385 JUN 2 9 2005 U.S. Nuclear Regulatory Commission Serial No.05-399 Attention: Document Control Desk MPS Lic/GJC RO Washington, DC 20555 Docket Nos. 50-245 50-336 50-423 License Nos. DPR-21 DPR-65 NPF-49 DOMINION NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION UNITS 1, 2 AND 3 10 CFR 50.59 AND COMMITMENT CHANGE REPORT FOR 2004 Pursuant to the provisions of 10 CFR 50.59(d)(2), the reports for changes made to the facility for Millstone Power Station Units 1, 2 and 3 (MPS 1, 2, and 3), are submitted via Attachments 1, 2 and 3, respectively. Attachment 4 reports changes made common to all Millstone Power Station units.
Additionally, during 2004, there were no commitment changes for MPS 1, 2, or 3. This constitutes the annual Commitment Change Report consistent with the Millstone Power Station's Regulatory Commitment Management Program.
If you have any questions or require additional information, please contact Mr. David W.
Dodson at (860) 447-1791, extension 2346.
Very truly yours, rice ice President - Millstone
Serial No.05-399 10 CFR 50.59 and Commitment Change Report for 2004 Page 2 of 2 Attachments: 4 Commitments made in this letter: None.
cc: U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406-1415 Mr. G. F. Wunder Project Manager U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 08-B-1A Rockville, MD 20852-2738 Mr. A. B. Wang Project Manager U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 7E1 Rockville, MD 20852-2738 Mr. R. Prince NRC Inspector U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406-1415 Mr. V. Nerses Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 8C2 Rockville, MD 20852-2738 Mr. S. M. Schneider NRC Senior Resident Inspector Millstone Power Station
Serial No.05-399 10 CFR 50.59 and Commitment Change Report for 2004 bc Page 1 of 1 bc: (*paper copies as noted; remainder electronic distribution)
MPS Standard Distribution W. J. Hayes K. A. Deveau M. B. Bennett (2 copies)*
Concurrence On S. E. Scace D. W. Dodson A. J. Jordan P. E. Grossman W. J. Hayes Verification of Accuracy
- 1. K. A. Deveau "E" mail memo to W. D. Bartron, dated 6/13/04, 50.59 Report 2004.
Action Plan/Commitments (Stated or Implied)
- 1. None
Serial No.05-399 10 CFR 50.59 and Commitment Change Report for 2004 bc Page 1 of 1 bc: (*paper copies as noted; remainder electronic distribution)
MPS Standard Distribution W. J. Hayes K. A. Deveau M. B. Bennett (2 copies)*
Concurrence S. E. Scace D. W. Dodson A. J. Jordan ct
- 0 O P. E. Grossman /
W. J. Hayes Verification of Accuracy
- 1. K. A. Deveau "E" mail memo to W. D. Bartron, dated 6/13/04, 50.59 Report 2004.
Action Plan/Commitments (Stated or Implied)
- 1. None
Serial No.05-399 10 CFR 50.59 and Commitment Change Report for 2004 bc Page 1 of 1 bc: (*paper copies as noted; remainder electronic distribution)
MPS Standard Distribution W. J. Hayes K. A. Deveau M. B. Bennett (2 copies)*
Concurrence S. E. Scace D. W. Dodson A. J. Jordan P. E. Grossman >,/ /D6 A, PA ,-/46W 1 A W. J. Hayes Verification of Accuracy
- 1. K. A. Deveau 'E" mail memo to W. D. Bartron, dated 6/13/04, 50.59 Report 2004.
Action Plan/Commitments (Stated or Implied)
- 1. None
Serial No.05-399 10 CFR 50.59 and Commitment Change Report for 2004 bc Page 1 of 1 bc: (*paper copies as noted; remainder electronic distribution)
MPS Standard Distribution W. J. Hayes K. A. Deveau M. B. Bennett (2 copies)*
Concurrence S. E. Scace D. W. Dodson A. J. Jordan P. E. Grossman W.J. Hayes L tk -
Verification of Accuracy
- 1. K. A. Deveau "E" mail memo to W. D. Bartron, dated 6/13/04, 50.59 Report 2004.
Action Plan/Commitments (Stated or Implied)
- 1. None
Attachment I 10 CFR 50.59 REPORT FOR 2004 Millstone Power Station Unit 1 Dominion Nuclear Connecticut, Inc. (DNC)
Serial No.05-399 10 CFR 50.59 and Commitment Change Report for 2004 Attachment 1/ Page 1 Manipulation of Refueling Gates and Plant Configuration Required to Support Transfer of Irradiated Hardware for the Reactor Vessel/Cavity Decommissioning Project - Millstone Unit 1 S1-EV-03-0001 Revision 0 SPROC OPS 02-2-021 Revision 0 FSARCR 03-MP1-1 Description This safety evaluation was written against activities which involved removing the gates between the Millstone Unit 1 spent fuel pool and reactor cavity, to support the transfer of irradiated hardware from the reactor vessel to the spent fuel pool for processing or storage. The Millstone Unit 1 Defueled Safety Analysis Report (DSAR) Section 3 was updated to indicate the gates between the spent fuel pool and reactor cavity are permanently closed and form part of the spent fuel pool boundary.
Reason Millstone Unit 1 is currently undergoing decommissioning and is in a defueled condition.
The permanently closed gate configuration supported the previous decommissioning plan to implement an Independent Spent Fuel Storage Installation (ISFSI) at Millstone Unit 1 which would transition the spent fuel from "wet" to "dry" storage. Once the fuel was offloaded from the spent fuel pool, the reactor vessel and cavity would be decommissioned and the water drained and processed from both the spent fuel pool and cavity. Since the decision has been made to maintain Millstone Unit 1 in uwet" storage for the duration of the SAFSTOR period and pursue draining down the reactor cavity and vessel, the gates were removed to support transfer of irradiated hardware from the reactor vessel to the spent fuel pool for processing. Therefore, the DSAR was revised to support gate removal. This DSAR change was implemented in 2004.
Summary There is no impact on the accidents previously evaluated in the DSAR which are defined as a family of fuel handling accidents. DSAR Section 5 defines the drop of a refueling gate as a design basis accident scenario. The DSAR statement in Section 3 that the gates are permanently installed did not eliminate the gate drop from the family of fuel handling accidents.
The consequences of a gate drop event had been previously evaluated in the DSAR and are well within 10 CFR 100 limits. This change remains within the bounds of previously evaluated malfunctions and load drop calculations. Therefore, the accident analyses for the family of fuel handling accidents previously evaluated is unaffected, and the radiological consequences of previously evaluated malfunctions remain the same.
Attachment 2 10 CFR 50.59 REPORT FOR 2004 MillstonelPower Station Unit 2 Dominion Nuclear Connecticut, Inc. (DNC)
Serial No.05-399 10 CFR 50.59 and Commitment Change Report for 2004 Attachment 2/ Page 1 Installation of Additional Liquid Radwaste Secondary Demineralizer T23B S2-EV-01-0024 Revision 0 DCR M2 01009 FSARCR 01-MP2-016 Description This design change modified the Clean Liquid Radwaste System by installing an additional Secondary Demineralizer Tank (T23B) and associated instrumentation to accommodate the new secondary demineralizer. The Final Safety Analysis Report (FSAR) Table 11.1-1 and Figure 11.B-1 were changed to reflect the increase in the number of demineralizers. In addition, Section 11.1.3.1 added information about the two Secondary Demineralizers.
Reason The addition of the secondary demineralizer improved demineralization of the coolant waste stream by increasing resin options, while improving processing flow rates.
Operating both secondary demineralizers in parallel doubles the processing flowrate while, at the same time, reduces curies discharged in this waste stream to the environment. Operating the two secondary demineralizers in series enables greater waste stream clean-up capability.
Summary The installation of the additional secondary demineralizer and associated monitoring instrumentation has no impact on the probability of occurrence of an accident or malfunction of equipment important to safety. Since off-site release paths are also unaffected, there is no impact on the consequences of an accident or malfunction of equipment important to safety. Because the design of the 'B" Secondary Demineralizer is similar to the previously installed "A" Secondary Demineralizer, no new failure modes are introduced by the proposed scope of this modification that creates the possibility of a new accident or malfunction of equipment important to safety.
Serial No.05-399 10 CFR 50.59 and Commitment Change Report for 2004 Attachment 2/ Page 2 Changes to the FSAR Chapter 14 Boron Dilution Accident Analysis and Technical Specification 3/4.1.1.3 and 3/4.9.8 Bases on Minimum RCS or Shutdown Cooling Flow(1)
S2-EV-02-0001 Revision 0 FSARCR 02-MP2-0003 LBDCR 2-2-02 Description Final Safety Analysis Report (FSAR) Section 14.4.6 was revised to incorporate changes made in the analysis of the boron dilution accident. The boron dilution accident was reanalyzed due to an error in the density of the unborated water that causes reduction in the Reactor Coolant System (RCS) boron concentration. The event was reanalyzed using the same methodology that was used in the previous FSAR analysis of record.
With the correction of the density of the unborated water causing the boron dilution, the minimum required shutdown cooling flow was increased.
The Bases for Technical Specifications (TS) 3/4.1.1.3 and 3/4.9.8 were modified to identify that the required minimum 1000 gallon per minute (gpm) flow is an analytical limit, and that plant operating procedures maintain the minimum flow at a higher value to account for instrument uncertainties.
Reason This change incorporated the results of a revised boron dilution accident analysis into the FSAR and modified the bases of Technical Specifications 3/4.1.1.3 and 3/4.9.8.
Summary No physical changes were made to plant equipment or structures as a result of this change. This change did not affect existing FSAR malfunctions or accidents nor create any new malfunctions or accidents.
The changes made to FSAR Section 14.4.6 incorporated the results of a revised boron dilution accident. The results of this revised analysis continue to meet the established acceptance criteria, providing adequate time for an operator to terminate the boron dilution prior to reaching criticality. No other previously evaluated FSAR accident is impacted by this change. Therefore, the change will not result in more than a minimal increase in dose consequences of an accident previously evaluated in the FSAR. The fuel cladding, RCS, and containment fission product barriers will not be challenged beyond that which was previously identified in the FSAR. Increasing the shutdown cooling flow operating band while in reduced inventory cannot cause an accident of a different type than previously evaluated, nor does it increase the likelihood or consequences of a loss of shutdown cooling flow or the consequences of a malfunction of any other SSC important to safety. As such, the change does not result inma design basis limit for a fission product barrier as described in the FSAR being exceeded or altered.
FSAR change was implemented in 2002, late submittal documented in CR-05-06523.
Serial No.05-399 10 CFR 50.59 and Commitment Change Report for 2004 Attachment 2/ Page 3 The methodology used in the revised boron dilution accident is the same as that used in the analysis of record. Therefore, the change did not result in a departure from a method of evaluation described in the FSAR.
Serial No.05-399 10 CFR 50.59 and Commitment Change Report for 2004 Attachment 2/ Page 4 Update to Millstone Unit 2 Design Basis Summary for Hydrogen Purge Startup Hydrogen Concentration S2-EV-02-0011 Revision 0 FSARCR 02-MP2-0013 DM2-00-0102-04 (DBS-2313C)
Description The Millstone Unit 2 Final Safety Analysis Report (FSAR) was updated in 2003 to address increased hydrogen generating material located in containment, which had not been addressed in the original design basis hydrogen generation analysis. The FSAR change decreased the post-accident containment hydrogen concentration for startup of the manual hydrogen purge system from 3% to 2.9%. This value is also contained in the Design Basis Summary (DBS) for Containment Hydrogen Control (DBS-2313C).
Reason The Unit 2 FSAR was changed to be consistent with the revised minimum concentration of hydrogen required for manual startup of the hydrogen purge system. This value was changed as a result of additional hydrogen generating material being added to containment during Refueling Outage 2R13. The design change for DBS 2313C, Table 8.1, is administrative only, updating DBS-2313C to ensure consistency between relevant design basis documents.
Summary The change to the post Loss of Coolant Accident hydrogen analysis allows more hydrogen generating material to be located in containment. The installed plant equipment is capable of preventing peak hydrogen concentration from exceeding the design limit even with a single failure. As a result of these conditions, the change to allow additional material to be in containment does not affect accidents or malfunctions previously evaluated in the FSAR, create a new type of event not previously evaluated in the FSAR, impact the fission product barriers as described in the FSAR, or require an evaluation under a different methodology from those previously evaluated in the FSAR.
Serial No.05-399 10 CFR 50.59 and Commitment Change Report for 2004 Attachment 2/ Page 5 Operability Determination: Low Pre-Charge Pressure Suspected in Charging Pump Pulsation Dampers Li1-B and L1l-CS2-EV-03-0007 Revision 0 MP2-063-04 Revision 0 Description On February 14, 2004, a new motor was installed on charging pump P18C. As part of the post modification testing, P18C was started with P18B running to monitor the starting pressures on the "C" discharge header and the common discharge header of the charging system. Test data on the UC" discharge header showed a pressure spike in excess of the given screening criterion. The test data, when compared to the readings taken after new pulsation dampeners were installed, showed a strong indication that the P18C pulsation dampener may have had less than the required pre-charge pressure. A previously written Technical Evaluation, M2-EV-03-0050, concluded that if a three-pump start were to occur with partially charged dampeners, discharge piping pressure could possibly become high enough to cause relief valves to lift and chatter.
Operability Determination (OD) MP2-063-04 concluded that disabling one pump would ensure that the relief valves would not be challenged upon a two pump simultaneous start. A two-pump start without the bladder type pulsation dampers was evaluated in a previously prepared Operability Determination, MP2-043-03. The compensatory measures for OD MP2-063-04, disabling one pump by maintaining the motor circuit breaker open during Modes 1, 2 and 3, when RCS pressure was greater than 1750 psia, were no different than those addressed in the previous OD. The SORC approved 50.59 evaluation associated with OD MP2-043-03 bounded the compensatory measures recommended for OD MP2-063-04. The pulsation dampener bladders have since been replaced and verified leak tight. This OD was then closed.
Reason This activity was performed to avoid challenging the charging pump relief valves following charging pump starts, thereby decreasing the likelihood of a failure in the charging system that could have resulted ina loss of charging flow.
Summary The compensatory measure identified in this OD resulted in the plant normally being operated with two charging pumps instead of three. Operating in this configuration would not increase the probability of an accident previously analyzed or create the possibility of an accident of a different type. Operating in this configuration would not have increased the dose consequences or challenges to the fission product barriers for any previously analyzed accident. Operating with two charging pumps instead of three would not increase the likelihood or consequences of a malfunction of a system, structure, or component important to safety nor create the possibility of a malfunction with a different result than previously analyzed. The Final Safety Analysis Report Chapter 14 accident analyses were unaffected by this change and remained bounding.
Serial No.05-399 10 CFR 50.59 and Commitment Change Report for 2004 Attachment 2/ Page 6 Replacement of MP2 Spent Fuel Pool Cask Crane Hoisting Equipment and Control System S2-EV-04-0001 Revision 0 DCR M2 02003 Description This Design Change upgrades the main and auxiliary hoists of the Millstone Unit 2 (MP2) spent fuel shipping cask crane to meet the requirements for a single failure proof crane in accordance with NUREG-0612 and NUREG-0554. These requirements ensure that the crane and the interfacing lift points are designed so that a single failure will not result in the loss of the capability of the system to safely retain lifted loads. This change increases the capacity of the main hoist to 125 tons; the auxiliary hoist capacity is maintained at 15 tons.
The control system for the 5 ton monorail hoist was relocated onto the new configuration without any additional modifications for single failure design. The controls for the 5 ton monorail hoist were modified only to allow for operation from the new control panels (both radio and hard wired backup control station). The monorail hoist was not upgraded to meet single failure proof requirements.
Reason The upgrade of the crane to meet the single failure proof requirements of NUREG-0554 and NUREG-0612 improves the reliability of the handling system through increased factors of safety and through redundancy or duality in certain active components. This upgrade, along with administrative and procedural controls for handling loads in a single failure proof configuration, permits qualified loads to be handled without the requirement to postulate load drop events. A load drop need not be postulated when it is handled in a single failure proof configuration.
Summary The upgrade of the crane to meet the requirements of single failure proof eliminates the need to perform drop analysis and the associated radiological consequences evaluation, since the intent of NUREG-0554 and NUREG-0612 single failure proof criteria is to reduce the likelihood of a crane failure. NUREG-0554 and NUREG-0612 requirements ensure that the crane and interfacing lift points are designed so that a malfunction of a component in the active load path will not result in the loss of capability of the system to safely retain lifted loads. The upgrade of the spent fuel cask crane will not increase the frequency of occurrence of an accident previously evaluated in the FSAR due to the conservatism of the design.
The increase in the capacity of the main hoist from 100 tons to 125 tons does not increase the frequency of occurrence of a Spent Fuel Cask Drop accident due to the design, fabrication and installation of the hoist as a single failure proof system in accordance with NUREG 0554 and NUREG 0612.
Serial No.05-399 10 CFR 50.59 and Commitment Change Report for 2004 Attachment 2/ Page 7 Millstone Unit 2 Compliance with 10 CFR 50.68(b)
S2-EV-04-0002 Revision 0 FSARCR 04-MP2-013 Description The Millstone Unit 2 Final Safety Analysis Report (FSAR) was updated to document Millstone Unit 2's compliance with 10 CFR 50.68(b) concerning Criticality Accident Requirements. This was done by adding 10 CFR 50.68(b) as a design criterion for fuel and reactor component handling equipment.
Reason As documented in Condition Report CR-04-06969, the U.S. Nuclear Regulatory Commission has stated they believe that 10 CFR 50.68(b) is part of the licensing basis for Millstone Unit 2, based on previous NRC reliance on the regulation for the Safety Evaluation Report for Millstone Unit 2 Technical Specification Amendment 274. As a result, documentation was put in place to demonstrate Millstone Unit 2's compliance with 10 CFR 50.68(b).
Summary For the most part, the requirements of 10 CFR 50.68(b) were the same as the now superseded design and licensing basis for the Millstone Unit 2 fuel storage and handling design criterion. The only exceptions were 1) 10 CFR 50.68(b)(5) requires the quantity of non-fuel Special Nuclear Material to be less than that required for a critical mass, and
- 2) 10 CFR 50.68(b)(1) requires that only unborated water be used in determining compliance with subcritcality for handling and storage of fuel assemblies. These additional requirements from 10 CFR 50.68(b)(1) and (b)(5) necessitated some procedure changes, but did not place the plant in any new conditions or configurations.
No new calculations needed to be performed, nor any physical plant modifications made in order to implement the procedure changes. These procedure changes do not affect the existing ESAR malfunctions or accidents, nor create any new malfunctions or accidents. Since K-effective of fuel continues to be maintained less than or equal to 0.95, there is no impact on the fission product barriers, nor are any adverse changes made to evaluation methods.
Attachment 3 10 CFR 50.59 REPORT FOR 2004 Millstone Power Station Unit 3 Dominion Nuclear Connecticut, Inc. (DNC)
Serial No.05-399 10 CFR 50.59 and Commitment Change Report for 2004 Attachment 3/ Page 1 Replacement of MP3 Spent Fuel Pool Cask Crane Hoisting Equipment and Control System S3-EV-03-0002 Revision 0 DCR M3-02004 Revision 0 FSARCR 03-MP3-28 Description Work continued on upgrading the main hoist of the Millstone Unit 3 Spent Fuel building indoor electric overhead trolley crane (3MHF-CRN1) to meet NUREG-0554 single failure proof design requirements. The modification includes installation of a new trolley on the existing runway. The new trolley includes new lifting and braking components and replacement of control systems for the main hook. The capacity of the main hoist remains unchanged at 125 tons. The physical load path for the crane was not changed, and travel remains limited from the cask pit area and into the spent fuel canopy building.
The Final Safety Analysis Report (FSAR) was revised to reflect the upgrade of the spent fuel shipping cask trolley to the single failure proof criteria.
Reason The upgrade of the crane to meet the single failure proof requirements of NUREG-0554 and NUREG-0612 improves the reliability of the handling system through increased factors of safety and through redundancy or duality in certain active components. This upgrade, along with administrative and procedural controls for handling loads in a single failure proof configuration, permits qualified loads to be handled without the requirement to postulate load drop events. A load drop need not be postulated when it is handled in a single failure proof configuration.
Summary The upgrade of the crane to meet the single failure proof criteria eliminates the need to perform drop analysis and the associated radiological consequences evaluation, since the intent of NUREG-0554 and NUREG-0612 single failure proof criteria is to reduce the likelihood of a crane failure. NUREG-0554 and NUREG-0612 requirements ensure that the crane and interfacing lift points are designed so that a malfunction of a component in the active load path will not result in the loss of capability of the system to safely retain lifted loads. The upgrade of the spent fuel cask crane will not increase the frequency of occurrence of an accident previously evaluated in the FSAR due to the conservatism of the design.
Serial No.05-399 10 CFR 50.59 and Commitment Change Report for 2004 Attachment 3/ Page 2 Millstone Unit 3 Cycle 10 Core Reload S3-EV-04-0001 Revision 1 DCR M3-03002 Revision 0 FSARCR 04-MP3-008 Description This Design Change Record (DCR) presented the documentation and references necessary to implement the Cycle 10 core reload and operate the Cycle 10 core safely and efficiently. The physical changes involved replacing 72 spent fuel assemblies with 72 feed assemblies labeled Region 12 (Batch M), of which 8 of those were Lead Test Assemblies (LTAs) of the Westinghouse Next Generation Fuel design. These LTAs will be placed in non-limiting core locations as required by Technical Specifications. The balance of the fresh Region 12 fuel will be of the RFA-2 design. The RFA-2 design has only one minor change in comparison to the once-burned Region 11 (Batch L) Robust Fuel Assembly design. The change is an overall increase in width of the mid-grid springs and local width of the dimple to provide increased margin to fretting wear.
This 50.59 evaluation also addressed the Final Safety Analysis Report Change Request (FSARCR) which updated the Millstone Unit 3 FSAR to include information regarding the Cycle 10 fuel and core, plus the Small Break LOCA (SBLOCA) reanalysis results from the new SBLOCA methodology recently approved by the NRC for Millstone Unit 3 application via Technical Specification Amendment 218.
Reason The changes in Cycle 10 fuel design were implemented to improve fuel product performance with respect to known issues such as spacer grid-to-fuel rod fretting wear, debris-related fuel rod fretting, top nozzle spring and screw loose parts, incomplete rod insertion, grid strap handling damage potential, hydrogen-generation rod internal pressure, and fuel rod clad and guide tube corrosion resistance.
The changes made by FSARCR 04-MP3-008 updated FSAR Chapters 4 and 15 to reflect the implementation of the Cycle 10 new fuel and core designs, as well as the implementation of the SBLOCA reanalysis results. These changes were necessary to update the FSAR to be consistent with the Cycle 10 core design and analyses of record.
Summary The implementation of the Millstone Unit 3 Cycle 10 reload core design did not affect any accidents or malfunctions evaluated in the FSAR, nor did it create a new type of event not previously evaluated in the FSAR. Implementation of the Cycle 10 reload core design did not create a negative impact on any fission product barrier as described in the FSAR.
The reload core design criteria and licensing basis acceptance criteria evaluations do not result in a departure from any evaluation methodology used in establishing the Millstone Unit 3 design basis or safety analysis.
Attachment 4 10 CFR 50.59 REPORT FOR 2004 Millstone Power Station Units 1, 2, & 3 Dominion Nuclear Connecticut, Inc. (DNC)
Serial No.04-399 10 CFR 50.59 and Commitment Change Report for 2004 Attachment 4/ Page 1 Common lonics Make-up Water Treatment Facility for Units 1,2, and 3 S2-EV-00-0009 Revision 1 DCR M2-00005 FSARCR 00-MP2-015 FSARCR 00-MP3-012 Description Permanent installation of the Ionics Water Treatment system was completed, providing a continuous supply of make-up water for the three Millstone units. The Millstone Units 2 and 3 Final Safety Analysis Reports (FSAR) were updated to reflect the new configuration.
Reason Installation of the Ionics Water Treatment system ensures a continuous supply of make-up water for the three Millstone units. With the installation of the permanent facility, total actual site make-up capacity remained the same, but the FSAR referenced site make-up water capacity was reduced from 600 gpm to 400 gpm. This corrects a long-standing discrepancy between the as-built versus designed water treatment facility at Millstone Unit 2 (CR M2-99-2829). Section 9.12 of the Unit 2 FSAR and section 9.2 of the Unit 3 FSAR were accordingly revised to reflect the new plant configuration.
Additionally, Section 9.12.4.2 of the Unit 2 FSAR stated that the pressure containing items were hydrostatically tested by the manufacturer at 1.5 times the design pressure.
The equipment for the common water header is supplied by the vendor, and all parts tested in accordance with ASME Section Vill. The next paragraph of the FSAR showed that all pressure vessels are tested in accordance with ASME Section Vil code requirements. Therefore, the statement regarding testing by the manufacturer was redundant, and consequently deleted.
Summary The function of the water treatment system was unchanged by this permanent modification. The total capacity of the common water treatment system remains the same as the original two vendor facilities located in Units 2 and 3. The new permanent system did not affect any equipment important to safety since the water treatment system does not affect the design of safety related systems. The safety related tanks that are filled by the water treatment facility have minimum levels which must be maintained in order to operate the plants. Failure to provide the required fill rate could result in not being able to maintain the required minimum Technical Specification volume, which would result in reaching a Limiting Condition for Operation. The consequences of failure to fill do not change as a result of these changes, therefore, there is no increase in the consequences of a malfunction, nor is there an increase in the probability of occurrence of a malfunction of equipment important to safety as previously evaluated.
There are no accidents associated with the installation of this common facility. The water treatment system does not initiate or prevent any of the accidents evaluated in the
-- Up Serial No.04-399 10 CFR 50.59 and Commitment Change Report for 2004 Attachment 4/ Page 2 FSAR. It does not interact with safety related equipment is such a way as to introduce an unanalyzed accident.