JAFP-99-0175, Annual Summary of Changes,Tests & Experiments for 1997/1998. with
| ML20195C146 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 06/02/1999 |
| From: | Michael Colomb POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| JAFP-99-0175, JAFP-99-175, NUDOCS 9906080024 | |
| Download: ML20195C146 (108) | |
Text
James A FitzPatr6c$ -
Nuclear Pcwor Plant 268 Laks Road ;
, i P.O. Box 41 J Lycoming, New York 13093 315-342-3640 E Michael J. Colomb l 4# Authon.ty sne e-we mce, !
June 2, 1999 j JAFP-99-01U j 1
United States Nuclear Regulatory Commission Attn: Document Control Desk Mail Station P1-137 Washington, D.C. 20555
Subject:
James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Annual Summary of Plant Changes, Tests, and Experiments for 1997/
1998 as Reauired by 10 CFR 50,59 Attachment 1: Annual Summary of Changes Tests, and Experiments for 1997/1998
Dear Sir:
i This letter transmits a summary of the changes, tests, and experiments implemented at the James A. FitzPatrick Nuclear Power Plant during 1997/1998 as required by 10 CFR 50.59(b)(2).
Provided with each summary is the Nuclear Safety Evaluation number (e.g. JAF-SE 0001), revision number, title, modification number, if applicable, followed by a brief description of the corresponding change, test, or experiment. '
If you have any questions concerning this report, please contact Mr. G. Tasick of my staff at (315) 349-6572. ;
Very truly yours, C l l
MICHAEL . COLOMB #j MJC:GJB:las cc: next page 200l ;
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9906090024 990602 I PDR ADOCK 05000333 R PDR l
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Regional Administrator U.S. Nuclear Regulatory Commission 475 Allendale Road
' King of Prussia, PA 19406 Office of the Resident inspector U.S. Nuclear Regulatory Commission P.O. Box 136 Lycoming, NY 13093 Mr. Joseph Williams, Project Manager Project Directorate i Division of Licensing Project Management U.S. Nuclear Regulatory Commission Mail Stop 8C2 Washington, D.C, 20555
I ATTACHMENT l Annual Summary of Channes, Tests, and Experiments for 1997/1998 i
introduction to the 1997/1998 Annual 10 CFR 50.59 Report i
10 CFR 50.59 (a)(1) states in part:
l The holder of a license...may (i) make changes in the facility as described in the safety analysis report, (ii) make changes in the procedures as described in the safety analysis report, and (iii) conduct tests or experiments not I described in the safety analysis report, without prior Commission approval, unless the proposed change, test or experiment involves a change in the technical specifications incorporated in the license or an unreviewed safety question.
10 CFR 50.59 (b)(2) states in part:
The licensee shall submit...a report containing a brief description of any changes, tests, and experiments, including a summary of the safety evaluation of each.
Unless otherwise noted, each safety evaluation listed concluded that it's subject change, test, or experiment did not:
e increase the probability of occurrence or the consequences of an accident or malfunction of structures, systems, or components important to safety previously identified in the FSAR:
Create the possibility of an accident of or malfunction of a different type than any preciously evaluated in the FSAR; e
Reduce the margin of safety as defined in the basis for technical specifications; And therefore, do not involve an unreviewed safety question as defined in 10 CFR 50.59.
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- 4 ATTACHMENT I Annual Summarv of Charm. Tests.' and F== iments for 1997/1998 I JAF-SE-91-060, REV.1: OSCILLOGRAPH PANEL REPLACEMENT MODIFICATION: F1-88-224 The purpose of this proposed activity is to install a new Digital Transient Recorder (DTR) in the plant Relay Room and restore the functionality of the old oscillograph. Power Generation (White Plains Office) has requested the replacement due to the outdated design, '
- unreliability and maintenance difficulties. The oscillograph panel will become a permanent termination point to extend all of the protective relay circuits to be monitored by the new DTR. This will allow the circuits to be extended without affecting the protective circuits.
These circuits are operating all the time, especially circuits on the 345KV and the 115KV system which provide power from and to the plant under existing design configuration.
With the plant in RO13 configuration, termination of these circuits can be made more advantageously and prevent plant trips.
JAFNPP conceptual design package, Modification No. F1-88-224, Revision 2, was prepared to initiate a modification to the plant. As a resu!t of the plant approval, JAFNPP Modification F1-88-224 was prepared to support the installation and documentation of the
! revised plant configuration.
, This modification does not alter the design operation or function of the plant protective relaying circuits. The existing oscillograph panel is OA Category ll/ill equipment; the new DTR equipment will remain QA Category ll/lli equipment.
The modification does not change or alter any protective relay scheme in the plant l . distribution system. Tl'.e DTR is only a monitoring device on the distribution scheme to evaluate plant trips caused by distribution problems in the plant or on the plant l transmission system. This data collected will be used by Power Generation to evaluate causes of line trips or possible plant trips.
. All of the cabling separation criteria wili be maintained in accordance with ECRIS, DCM-25A, FSAR criteria JAF-RPT-ELEC-02075. This modification cannot interact with any safety system to affect plant operation. No new malfunction of a different type other than
- j. any evaluated previously in the FSAR can occur. Section 3.9 of the technical specifications has been reviewed for possible interactions and it has been determined that there are none.
The physical changes made to the plant were evaluated and have demonstrated that these l changes will not have an adverse impact on the plant. This modification will not impact radiological / environmental issues or produce a release to the environment and maintains the FSAR analysis.
The proposed activity evaluated in this Nuclear Safety Evaluation does not result in an unreviewed safety question.
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ATTACHMENT I Annual Summary of Channes. Tests. and Emeriments for 1997/1998 JAF-SE-91-061, REV. 5: REPLACEMENT OF CONTROL SWITCHES FOR 10MOV-12A/B MODIFICATION: M191-156 JAF modification M1-91-056 replaced the existing keylock open/ closed Control Room switch on the 09-3 panel with a three pcsition open/stop/ closed switch for the RHR Heat
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Exchanger outlet valve (10MOV-12A/B). This allows remote throttling of 10MOV-12A/B '
from the Control Room.
Following implementation of this modification, the RHR System has been operated as follows: the reactor coolant water cooldown rate during shutdown cooling operations has been controlled by throttling either 10MOV-12A/B or 10MOV-66A/B to shunt a percentage of flow through the RHR Heat Exchangers E2A/B The outboard LPCIinjection valve (10MOV-27A/B) is roly throttled to control RHR System flowrate during LPCI operations.
The applicable portions of the j4F FSAR and Technical Specifications have been reviewed and no unreviewed safety question or change to the Technical Specifications result from this modification.
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ATTACHMENT l Annual Summarv of Channes. Tests, and Experiments for 1997/1998 l
JAF-SE-91-095, REV. 2: SUBSTITUTION OF THE AUXILIARY BOILER STEAM SUPPLY MODIFICATION: N/A l This safety evaluation demonstrates that the unavailability of the non-safety related Auxiliary Boiler System and removal and substitution of the steam supply of this system with other methods of heating are acceptable.
This safety evaluation presents the evaluation of the impact of the removal and l substitution of the Auxiliary Boiler steam supply upon the plant safety. This evaluation addresses the design bases of the various plant uses of the Auxiliary Boiler steam supply, the safety significance of each of these uses, their substitutions with other heating sources, and the effects of these substitutions on overall plant safety.
The Auxiliary Boiler System modified heat source does not affect the safety of the plant nor is it required to mitigate the consequences of any plant accident evaluated in the FSAR or Safety Evaluation Report. And, since the proposed activity neither increases the probability or consequence of an accident, an unreviewed safety question does not exist.
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ATTACHMENT I Annual Summary of Channes, Tests, and Experiments for 1997/1998 [[::JAF-SE-92|JAF-SE-92]] 246, REV.1: WELD EXAMINATION AND POST-WELD HEAT TREATMENT REQUIREMENTS MODIFICATION: N/A i This Nuclear Safety Evaluation addresses changes to the NDE requirements for Q-1, Q 2, and Q-3 Pressure-retaining Piping Welds that were used to satisfy the original construction code following RRM activities. NDE in excess of the construction code requirements was imposed during construction. The revised NDE requirements are based on ASME B31.1.0, 1967 Edition with Addenda A (1969) and March 1969 Erratum (the original construction code) and applicable Code Cases for reactor coolant pressure boundary piping.
There is no change in the NDE requirements for Q-1 piping. For Q-2 and Q-3 piping the NDE requirements were changed to be consistent with the design code. The 20% random radiographic testing and magnetic particle / liquid penetrant testing performed on these systems imposed during construction is eliminated. These requirements are satisfied during implementation of the ASME section XI program during RRM activities and place the emphasis for NDE on piping important to plant safety. The section XI program has been expanded to include the requirements of later, approved ASME code editions, addenda and code cases as far as practicable.
The Post Weld Heat treatment requirement for P-l (carbon steel) materials was changed from ASME B31.1.0 -67 requirements to the requirements of ASME section lil,1992 edition.
Compared to the original construction code (B31.1.0) the ASME section lli code,1992 edition, is a more refined code for PWHT. The change from the original construction code to section ill for PWHT does not reduce the margin of safety.
After reviewing the FSAR and Technical Specification it is concluded that changes to the l NDE program do not involve an unreviewed safety question. The revised NDE includes l various methods such as surface examinations (visual, magnetic particle, liquid penetrant) l and volumetric (radiographic and ultrasonic). The changes in the NDE program are commensurate with the original construction code and will provide assurance that the structural integrity of Q-1, 2 and 3 piping will be maintained at the level required by the original acceptance standards, 1
l The NDE changes do not change the design or prevent any system from performing its I function and the probability of an accident or malfunction is not increased. The change l does not create any new accident or equipment malfunction not previously evaluated in the
! FSAR.
All future examinations, except visual, will be conducted using methods and techniques of ASME section V and XI that meet the ASME code requirements of the current ISI program.
Later codes may be used and defined in the nuclear safety evaluation for the modification. ;
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ATTACHMENT I Annual Summary of Chances. Ients. and Experiments for 1997/1998 l
JAF-SE-93-017, REV. 5: CO2 VENTING IN RELAY ROOM AND HVAC CONTROL l.OGIC l MODIFICATION: F1-92-377 This modification will upgrade the Relay Room CO2 system vent path to improve the required CO2 concentration and to isolate the Relay Room ventilation system to preclude CO2 leakage into the Control Room via the ventilation system. Also, this modification will limit intrusion of CO2 vapor into the Control Room space by replacing the Relay Room / Control Room communicating door 76FDR-A-300-16 with a new pressure rated low leakage door and by repairing leaks found during fan pressure testing through tN Control Room electrical penetrations and structural components. This modification 41 also improve the existing Relay Room CO2 Discharge System control equipment so as to minimize the effect of postulated design basis single failure occurrences. This modification will not increase the probability of occurrences or consequence of an accident or malfunction of structures, systems, of components important to safety previously evaluated in the FSAR. In addition, this modification to the Relay Room CO2 ventilation system and fire door will not create an accident or malfunction of a different type than evaluated in the FSAR. This modification will demonstrate that the Relay Room CO2 system will meet or exceed its design concentration requirements per NFPA 12, Reference 9, by alternate testing methodologies in lieu of a full discharge test. The system concentration will be demonstrated by tracer gas testing (Special Test Procedure STP-76AU and Nuclear Safety Evaluation Number JAF-SE-95-004) and by Calculation Number JAF-FPT-FPS-02009, as described in Alternate Testing Methods letter. Control Room habitability will be determined based on a CO2 short term exposure limit of three (3) percent by volume in the Control Room. The three (3) percent by volume is acceptable for {
short term exposure as evaluated. This concentration will be used in conjunction with Relay Room fan pressure leakage testing to demonstrate that Control Room habitability is not compromised during a Relay Room CO2 discharge.
l The modification to the Relay Room ventilation, CO2 systems, and fire door does not reduce the margin of safety as defined in the basis for any Technical Specification. The modification to the Relay Room CO2, Ventilation system and Fire Door is being made to the l same design standards as the existing systems. The single failure modifications will l
minimize the effect of the single failure concerns in the existing system. l The modifications do not involve changes to the Technical Specification. Fire Protection licensing commitments for the J.A. Fitzpatrick Nuclear Power Plant are no longer contained within the Plant Technical Specification document. Administrative Procedure AP-01.04 ]
contains commitment information previously present within the JAF Tech. Spec. This information is not affected by this modification. This modification is being done to return j the Relay Room CO2 system to operable status.
l The modification will not affect the environmental impact of the plant or involve an unreviewed safety question. Following activation of the CO2 system the same quantity of CO2 will be purged from the Relay Room and vented to the outside atmosphere as is presently the case. This modification does, however, help to minimize CO2 leakage into other areas of the plant which will ensure a safe plant environment.
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l ATTACHMENT I Annual Summary of Channes. Tests, and Experipents for 1997/1998 I
I JAF-SE-93-025, REV 0: REVERSE OSMOSIS SYSTEM FOR THE MAKE-UP i DEMINERALIZERS (MUD) SYSTEM MODIFICATION: F191-149 Modification F1-91-149 will install a 60 gpm Reverse Osmosis (RO) unit and make l
enhancements to the Make-Up Demineralizer (MUD) system logic. Revisions to plant drawings and the FSAR are necessary to incorporate this modification. The implementation !
of modification F1-91-149 does not change the design bases of the FSAR. Furthermore, i the MUD system's ability to provide an adequate supply of treated water for plant '
operating requirements will not change. The MUD system performs no safety significant function and has no interface with any component or system credited in the safety analysis for accident mitigation.
These changes do not involve an unreviewed safety question or an unreviewed environmental question.
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ATTACHMENT I Annual Summary of Channes. Tests. and Experiments for 1997/1998 l JAF-SE 94-038, REV. 3: TEMPORARY POWER REQUIREMdNTS DURING BUS OUTAGES PER OP-46A MODIFICATION: N/A The purpose of this Nuclear Safety Evaluation is to evaluate the acceptability of installing temporary power feeds as identified in a proposed revision to Maintenance Procedure MP-071.17, rev. 6. These feeds are required to maintain selected station loads energized on the 11500,12500,11600, and 12600 safety related busses during outages.
Two different bus "line-up scenarios" will result from these temporary power feeds. A non-safety r; lated bus will supply power to a safety-related bus (or loads) or a safety-related bus on one train will supply power to a safety-related load on the other train.
The evaluation credits the procedural and design requirements that:
a) Loads / buses receiving temporary power will be isolated from their normal source of power by racking out their respective feeder breakers (no back feeds),
b) No credit is taken for the loads receiving ternporary power. These loads are considered to be inoperable.
c) Adequate overcurrent protection and coordination has been established for all temporary power feed sources therefore no single fault in any safety-related load will adversely affect any of the temporary power sources, d) The load capacity margin for the temporary sources identified in MP-071.17 has been evaluated for the plant conditions applicable far this procedure and adequate load capacity is available.
The evaluation determined that no unreviewed safety question exists due to the fact that the electrical distribution system configuration, with the temporary power installation, does not affect overall system performance in a manner outside of minimum Cold Shutdown Technical Specification equipment operability requirements. For the duration of the temporary power feed installation, the powered loads are still considered administratively inoperative. The temporary powered loads will not play a role in mitigating radiological consequences. The installation of the temporary power sources does not introduce any new accident initiators or failures beyond that previously analyzed by the SAR. The potential failure modes introduced by the circuit modifications is bounded by the SAR, section 14.5.5.4, accident analysis for the loss of power to auxiliaries during power operation.
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ATTACHMENT I Annual Summary of Channes. Tests, and Experiments for 1997/1998
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l JAF-SE 94-069, REV. 2: ONSITE INTERIM WASTE STORAGE FACILITY MODIFICATION: F1-82-026 The FitzPatrick interim waste storage facility was constructed to temporarily store low-level radioactive wastes (including resins, sludges, filters, and dry-active wastes and radioactive trash) produced by plant operations. The facility was evaluated for its ability to minimize impacts to plant personnel and the offsite environment and to comply with federal guidelines when storing radioactive waste as well as non-waste plant components or equipment. This is accomplished through a combination of waste and container j characteristics, container dose rate limits, and administrative measures.
The wastes for storage will be placed in containers that meet the current requirements for disposal and other components and equipment placed in storage will meet the same general conditions. The maximum average contact dose rate for containers in the resin storage area is 50 R/ hour, for containers in the DAW Storage Bay is 125 mR/ hour, and for containers in the below-grade DAW Compartment is 1 R/ hour. The waste containers, and other components or equipment that are stored in the facility, will be subject to the radiological surveillance program, an inspection and sampling program, a fire detection system, and procedural controls which include controls on dose rates, total curies, and fire loading to remain within the limits of this evaluation.
Based on the above analysis, the use of the interim waste storage facility for the storage of plant radioactive waste, including empty containers or plant components and equipment within the limits as described herein, does not constitute an unreviewed safety question.
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ATTACHMENT I Annual Summarv of Cham . Tests. and FHments for 1997/1998 l-
'JAF-SE-94-108, REV. 3: FIRE BARRIER PENETRATION SEALS, FIRE DAMPERS AND I r FIRE DOORS, AP-01.04 CHANGES MODIFICATION: N/A The purpose of this Nuclear Safety Evaluation (NSE) is to evaluate any safety concerns
' associated with changes to Administrative Procedure AP-01.04, Tech Spec Related Requirements, Lists, and Tables *. Revision O of this NSE changed the inspection interval from once an operating cycle of 18 months to an operating cycle of 24 months for accessible fire barrier penetration seals, fire dampers and fire doors. Revision O of this NSE also changed the inspection practice from 100% of the total population once per operating.
. cycle to a statistically based sample.
Revision 1 of this NSE changed the inspection practice from a statistical based sample of fire barrier penetration seals and fire' doors once per operating cycle (i.e.18 months) to a sampling of 10% of fire barrier penetration seals, fire dampers and fire doors once per operating cycle (i.e. 24 months) with 100% of the total population inspected within 10 l operating cycles.
Revision 1 also deferred inspection of inaccessible fire barrier penetration seats (inspected from a single side only) for the 1994 surveillance inspection and deferred inspection of l inaccessible fire barrier penetration seals for future surveillance inspections until they become accessible or inspect them within 10 operating cycles.
Revision 1 also changed the compensatory measures, i.e. fire watch requirements associated with inoperable fire barrier penetration' seals, fire dampers and fire doors and clarified the intent of AP-01.04 Technical Requirement 1.1.2.F and Surveillance l Requirement 1.2.2.F.
1 l 1 i Revision 2 of this NSE revised the surveillance frequency for fire dampers to once per operation cycle and discontinued the inspection of inaccessible penetration seals.
Revision 3 of this NSE deletes the requirement to perform a "GL 86-10 evaluation" prior to discontinuing inspection of inaccessible penetrations.
Originally, the Technical Requirements for fire barrier penetration seals, fire dampers and i i
fire doors required 100% visualinspecton of these components each plant opert:ing cycle j (18 months plus 4.5 month grace period). The basis for this requirement was to minimize ,
the possibility of a single fire rapidly involving several areas of the facility prior to detection ;
and extinguishment.
i The reduced scope and increased inspection interval will not adversely affect the ability of the fire barriers to perform their design function. This is because sampling of the penetration seal and fire door population verifies that the administrative controls associated with penetration seals are adequate to prevent unintended changes to the installed .
- configuration and the preventative maintenance program is effective in ensuring that fire l doors are maintained properly. In addition, the increased inspection interval for fire dampers will not result in any safety concerns due to the improvements made to the dampers. ,
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ATTACHMENT I Annual Summary of Channes, Tests, and Experiments for 1997/1998 JAF-SE 94-139, REV. 2: EVALUATION OF MG SET ROOM PLUGS REMOVAL FOR ACCESS TO STEAM TUNNEL DURING PLANT OPERAITON - ~
" ACTIVITY CANCELLED" I MODIFICATION: N/A The pt,rpose of this Nuclear Safety Evaluation is to evaluate the significance of removing one or more of the Steam Tunnel removable concrete floor plugs in the Motor Generator (MG Set) Room during plant operation and to determine if there are any unreviewed safety questions.
Revision 2 of this safety evaluation cancels the originally proposed activity.
Deviation / Event Report (DER) 96-0968 and INPO OE 7946 addressed an industry event that determined removal of individual room floor / ceiling plugs could potentially affect existing Steam Leak Detection (SLD) equipment. The JAF review of OE 7946 and the response provided to DER-96-0968 concluded that removal of the MG Set Room floor plugs for access to the steam tunnel might adversely affect the steam tunnel leak detection j equipment. An engineering evaluation to evaluate if removal of the plugs has an adverse !
effect on the SLD system is not practical. Therefore, based on the NYPA review performed i for DER-96-0968, removal o the MG Set Room floor plugs to allow access to the steam tunnel during plant operation is not allowed by this nuclear safety evaluation.
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ATTACHMENT I Annual Summary of Channes. Tests, and Experiments for 1997/1998 JAF SE 95-013, REV,1: INSTALLATION OF BLANK FLANGE IN PLACE OF RUPTURE ,
DISC PARALLEL TO 33RV-101 '
MODIFICATION: TEMPORARY MODIFICATION 97-045 ,
The pu6 pose of this Nuclear Safety Evaluation is to evaluate the acceptability of installing a l blank flange in place of a rupture disc which along with Relief Valve 33RV-101, provides I overpressure protection to Mixed Resin Storage Tank 33TK-15.
Replacing the rupture disc installed in parallel to 33RV-101 is acceptable because, per JAF-CALC-CND-02014,33RV-101 has sufficient capacity to provide over pressurization protection for 33TK-15. The setpoint for 33RV-101 is 75 psi which corresponds to the i maximum design pressure for 33TK-15. Any com'ns relative to plugging of the relief l valve operator have been eliminated per Mod. D l-95-U33 which replaced the operator with a design properly suited for the installed environment. Also, the relief valve is of a " soft ,
seat" design which assists in maintaining a seal with resin potentially in the seat.
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The proposed activity evaluated in this nuclear safety evaluation does not result in an unreviewed safety question.
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ATTACHMENT I Annual Summary of Channes Tests, and F== iments for 1997/1998 JAF-SE 95-034, REV.1: EVALUATION OF REMOVAL OF VARIOUS CONTAINMENT ISOLATION VALVES FROM AP-01.04 MODIFICATION: N/A s
Several valves are currently listed in Administrative Procedure AP-01.04, " Tech Spec Related Requirements, Lists, and Tables" as primary containment isolation valves, although 1
they do not perform a containment isolation function. This Nuclear Safety Evaluation will document the reasons for their removal from the procedure, as well as describe the {
necessary FSAR, IST Program and procedure changes.
The removal of 15RBC-21 A/B, -22A/A, 24A/B, -26A/B, -33, 46 ESC-15A/B, 46ESW-16A/B,13MOV 39 and 23MOV-57 from the list of containment isolation valves listed in I AP-01.04 does not result in any unreviewed safety question. Primary containment isolation '
is provided by other valves in the associated lines and all applicable design criteria in 10CFR50 Appendix A are met without these valves.
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' ATTACHMENT I Annual Summary of Channes Tests, and Experiments for 1997/1998 JAF SE-95-044, REV.1: JUSTlFICATION TO SECURE THE FUEL POOL COOLING AND CLEAN-UP SYSTEM (FPCCS) TO SUPPORT MAINTENANCE ACTIVITIES MODIFICATION: N/A The purpose of this safety evaluation is to address the acesptability of securing the Fuel Pool Cooling and Clean-up System (FPCCS) (per revised section F of OP-30) to perform maintenance activities which require the system to be shut down. This safety evaluation covers each time the FPCCS is required to be shut down for maintenance when the fuel pool heat-up rate is known to be less than 1'F/hr. Limitations will be established to ensure fuel pool temperatures stay below design temperatures.
A safety evaluation is required to secure the FPCCS due to the requirement in FSAR sec.
9.4.4 stating that the system is normally in operation while fuelis stored in the Spent Fuel Pool.
Revision 0 of this NSE was written to allow a FPCCS shutdown for cycle 12 only. The NSE is being revised to allow a system shut down as long as the fuel pool heat-up rate is known to be less than 1 *F/hr.
Securing the FPCCS will deviate from the FSAR relative only to the descriptive sections (i.e., discussions of normal system operation) and will not impact system safety. Adequate j precautions are identified in the operating procedure (OP-30) to alert the operators of excessive fuel pt,ol temperatures. FPCCS can be restored using OP-30, as needed, to maintain temperatures below spent fuel pool design temperatures.
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l Annual Summary of Channes Tests. and Experiments for 1997/1998 l I
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'JAF SE-95-046, REV. 3: TECH SPEC RELATED REQUIREMENTS, LISTS, AND TABLES, 1
1 ATTACHMENT 2, FIRE PROTECTION CHANGES MODIFICATION: N/A The purpose of.this Nuclear Safety Evaluation (NSE) is to evaluate any safety concerns associated with changes to Administrative Procedure AP-01.04, " Tech Spec Related
! Requirements, Lists and Tables"."
Revision 3 makes changes to the Emergency Battery Lighting Surveillance Requirements that were added as a result of revision 1 to this safety evaluation. The Emergency Battery :
' Lighting (EBL) battery discharge testing frequency will be decreased from every 18 months to every 36 months with battery replacement every 5 years. This change is based on engineering report JAF-RPT FPS-02699.
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Revision 3 also changes the frequency for the Surveillance Requirement,1.2.2.A.1.e.4, for
' the High Pressure Water Fire Protection System Functional Test for the low pressure start ,
! switches that automatically actuate the electric and diesel driven fire pumps from Once/18 !
months to Once/24 months. This change is based on JAF-RPT-MISC-02082, Revision 1, !
" Miscellaneous Systems Surveillance Test Extensions," dated October 1995.
l In addition, 76ELB-SW-260-6 is not required to support manual actions to achieve hot shutdown (Appendix R) (Reference 93). This light illuminates the fire pump controller panel 76FPP-4 for the east diesel fire pump (76P-4). This EBL unit was removed from the table which identifies the required Appendix R EBLs (section 2, item 15 of this NSE).
l The proposed activity evaluated in this nuclear safety evaluation does not result in an unreviewed safety question.
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I ATTACHMENT I Annual Summary of Chanaes. Tests, and Experiments for 1997/1998 l
JAF-SE 96-011, REV. 4: MOVING FUEL WHILE MSIVs ARE BEING WORKED WITH j MAIN STEAM LINE PLUGS INSTALLED I MODIFICATION: N/A The purpose of this nuclear safety evaluation is to evaluate the acceptability of performing I maintenance that would breach the pressure boundary of Main Steam (MS) Line i )
Components during refueling with main steam line plugs in use. FSAR Figure 9.17-1 specifically addresses working the SRVs after the MS Line Plugs are installed; however, other inboard steam isolation valves are not specifically addressed. The MS Line Plugs are used as a reactor vessel boundary during maintenance to MS line components (MSIVs, RCIC inboard Steam Valve, HPCl inboard Steam Valve, 29MOV-74, SRVs, and !
miscellaneous steam drain valves located upstream of inboard steam isolation valve) which l
would normally provide this function during refueling operations. This evolution would only ;
be required if the MS line component maintenance were to breach the pressure boundary.
The evaluation also addresses the initial plant condition where the Torus would be '
unavailable as a supply of water to ECCS systems.
During normal refueling operations, the main steam line plugs provide a reactor vessel boundary that prevents refueling water from draining through the Main Steam lines during maintenance on MS Line components. In the event of MS line component maintenance affecting the pressure boundary, the main steam line plug will provide the only barrier preventing leakage. This plug will then become the boundary and allow refueling operations to continue concurrent with MS Line component maintenance.
The proposed activity evaluated in this nuclear safety evaluation does not result in an !
unreviewed safety question.
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r ATTACHMENT I Annual Summary of Chanaes. Tests, and EXDeriments IOr 1997/1998 JAF-SE-96-013, REV.1: REFUEL BRIDGE UPGRADES MODIFICATION: M195-122 The purpose of the Refuel Bridge Upgrades (Minor Modification Ml-95-122) is to increase reliability, improve equipment operating speed, enhance efficiency of fuel handling operations, and generally improve the performance of the Refuel Bridge to reduce the time required for refueling the reactor.
The scope of this modification consists of the addition of several new Refuel Bridge Syster.1 components, the replacement of some existing components, and the addition of new interlock and control features to upgrade the Refuel Bridge. The purpose and function of the Refuel Bridge System remain unchanged.
The purpose of the complete or partial turnover of the modification is to allow Operations to use the Refuel Bridge within the bounds of the modification turnover document limitations.
A 10CFR50.59 Nuclear Safety Evaluation has been performed for the Minor Modification M1-95-122 and a partial turnover in accordance with NYPA Procedure MCM-4 (Nuclear Safety And Environmental Impact Screens And Nuclear Safety Evaluations). This evaluation has considered the scope of the modification and any potential effects that the modification might have on Structures, Systems, or Components (SSCs) that could be affected by the modification. A detailed review of the FSAR sections and Technical Specifications applicable to potentially affected SSCs has been performed. As a result of this review, it is concluded that no unreviewed safety questions exist regarding the implementation of the Minor Modification M1-95-122 or the complete or partial turnover, l
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ATTACHMENT I Annual Summary of Channes Tests. and Experiments for 1997/1998 JAF SE 96-016, REV. 3: ACCEPTABILITY OF ESTABLISHING A ONE HOUR POST ACCIDENT OPERATING TIME REQUIREMENT FOR THE ADS SYSTEM TO MITIGATE LOCA MODIFICATION: N/A The purpose of this Nuclear Safety Evaluation (NSE) is to evaluate the acceptability of establishing a one hour post accident operating time requirement for the ADS system to mitigate the consequences of a LOCA with a pipe break cross-sectional areas 2. 0.5 ft 8.
This qualification requirement is required because Safety Relief Valve Solenoid Operated Valves (SRV SOVs) which are being installed under Type 1 Change D1-96-007 have been qualified by the vendor to lower radiation exposure values than pre D1-96-007 SRV SOV's.
This NSE will also examine the post D1-96-007 plant configuration to ensure this Type 1 Change has no impact on plant safety or availability. This examination is desirable due to complexities in installation details which came to light during the installation of D1-96-OO7.
The purpose of revision 3 is to state that the electrical connectors will be installed in flex conduit and spliced in a condulet.
This NSE concluded that the above changes do not involve an unreviewed safety question.
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I ATTACHMENT I l Annual Summary of Channes. Tests. and Exneriments for 1997/1998 l l
l JAF SE-96-017, REV. 3: SINC'2 CONTAINMENT ISOLATION VALVE AND CLOSED l
LOOP OUTSIDE CONTAINMENT MODIFICATION: N/A This safety evaluation assesses whether an unreviewed safety question exists if the requirement for Appendix J, Type C testing of certain Containment isolation Valses (CIVs) is deleted. The original NSE and revision 1 of this evaluation had deleted the cbssification of these valves as CIVs while retaining all mechanical functions. Revision 2 restored the classification of these valves as Containment isolation Valves in Table 7.3-1 and AP-01.04, but clarified that they do not require Appendix J testing. Revision 3 clarifies the FSAR sections needing revision.
This change is based on the use of a different containment isolation configuration, i.e. a single ClV and a closed loop outside containment. The use of such configurations is 2 permitted in GDC 55 and 56 provided they are demonstrated to be acceptable on some other defined basis. The basis for such configurations being acceptable is described in i ANS-56.2, which was endorsed by Reg. Guide 1.141.
This proposal makes no physical change to the f acility as described in the SAR. The only change is a reclassification of several valves as not requiring Appendix J testing in FSAR Table 7.31 and AP-01.04. The justification for this change is consistent with GDC 55 and
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No Unreviewed Safety Questions exists since removing Appendix J Type C testing for !
these valves does not increase the probability or consequences of an accident previously l evaluated in the SAR, nor is the probability or consequences of a malfunction of equipment I i
important to safety previously evaluated in the SAR increased. The possibility of an accident or malfunction of a different type than any previously evaluated in the SAR is not i created and the margin of safety as defined in the basis for any Technical Specification is not reduced.
A Technical Specification change is not required because the CIVs subject to 10CFR50 Appendix J Type C testing are not listed in Technical Specifications.
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l ATTACHMENT I Annual Summary of Chanaes, Tests, and Experiments fo- 1997/1998 i
JAF SE 96-028, REV.1: INSTALLATION OF HOT WATER BOILER 87HWB-1 A '
MODIFICATION: F195-109 This modification, which provides an alternate means of supplying hot water to the plant heating system, will not impact any safety-related or environmentally qualified components or systems or affect overall plant safety.
The purpose of this modification is to install a hot water boiler to supply back-up hot water to the non-safety related plant heating system.
This modification includes a packaged boiler, associated piping and required electrical supply to the hot water boiler. This hot water boiler will be located in space vacated by removal of Auxiliary Boiler 87AB-1 A The design and installation complies with applicable I codes, standards, p! ant criteria and environmental requirements.
On the basis of these evaluations, it is concluded that the change to the plant as a result of this modification does not represent an unreviewed safety question.
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ATTACHMENT I l Annual Summary of Chana=s, Tests, and Evaeriments for 1997/1998 JAF-SE-96-037, REV.1: ISP-1 ACCEPTANCE CRITERIA l MODIFICATION: N/A i
l This safety evaluation proposes to change leakage rate acceptance criteria for Excess Flow l Check Valve testing because instrument line breaks are enveloped by previously analyzed transients and accidents in the SAR. The instrument line 1/4 in. orifice is credited in the l SAR for reducing the blowdown in the event of a break in a 1 inch section of instrument l
pipe outside the primary containment to a rate that will not result in overpressurizing the secondary containment. The analysis of the radiological consequences of the Main Steam Line Break Accident bounds the radiological consequences of an instrument line break.
Removal of the qualitative test from the acceptance criteria does not lower the safety ,
threshold since the quantitative acceptance criteria remains to verify valve operability I which is in accordance with accepted testing practices. The qualitative test is also not required by the ASME Code. The safety evaluation concludes that the change of acceptance criteria does not constitute an unreviewed safety question pursuant to 10CFR50.59.
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1 ATTACHMENT I Annual Summary of Channes. Teats, and Experiments for 1997/1998 JAF-SE 96-042, REV,5: USE OF THE DECAY HEAT REMOVAL SYSTEM IN VARIOUS PLANT MODES AND CONFIGURATIONS MODIFICATION: F1-95-121 A Decay Heat Removal (DHR) system is provided which is completely independent of the FPC and RHR systems and b powered independently of the in-plant electric distribution system. The DHR is a non-safety related system with a design maximum heat removal capacity of 45 X 10' BTU /HR at a wet bulb temperature of 73*F. The DHR system has a .
normal heat removal configuration which provides two loops, either of which can remove 30 X 10', BTU /HR at a wet bulb temperature of 73*F. The DHR system is designed to supplement or to substitute for either the RHR system and/or the FPC system to provide additional operating flexibility. The DHR system can remove decay heat from fuel elements located in the SFP with the fuel pool gates in place. The DHR system can also remove decay heat from fuel elements in the reactor pressure vessel (RPV) when the refueling canalis flooded and the SFP gates are removed.
The discussions and evaluations provided above provide the bases for the conclusion that the implementation and operation of the DHR system does not constitute an unreviewed safety question.
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ATTACHMENT I Annual Summary of Channes, Tests, and Exoeriments for 1997/1998 l
JAF-SE 96-045, REV,2: POWER UPRATE STARTUP TEST PLAN MODIFICATION: M1-96-061
- This evaluation documents review of the safety issues associated with the power uprate startup test program.
j The power uprate startup test plan will be implemented during startup from Refueling Outage 12. The scope of testing to be performed is documented in an approved engineering report and has been found acceptable. The individual procedures to be used to implement the test plan have been reviewed and found acceptable. The implementation of the power uprate test plan to meet a license condition to follow a startup testing program as described in GE Licensing Topical Report NEDC-31897P-A does not constitute an unreviewed safety question as defined in 10CFR50.59. The test requirements specified ensure that there are no unreviewed safety questions associated with the use of the LEFMs to measure feedwater flow and determine core thermal pawer, i
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Annual Summary of Char- . Tests. and F=aa iments for 1997/1998 L
. JAF-SE 96-048, REV. 2: REVISION TO FSAR TO RAISE MAXIMUM ALLOWABLE LAKE TEMPERATURE FROM 82 F TO 85'F l
l MODIFICATION: F1-97-016 This rafety evaluation supports plant operation for current rated core thermal power of 2536 MWT with a lake temperature design limit of 85 F at JAFNPP. Specifically, the impact on the performance of the RHRSW and ESW systems was evaluated. This safety -
evaluation is based on an examination of the impact of the higher service water -
temperature on SRV operation, LOCA related containment responses, ECCS performance during design basis events, the operabiltty of the RHRSW pumps, ECCS performance for l
degraded events, and safety-related cooling systems utilizing ESW as a heat sink. It is concluded that the 85'F service water temperature does not have an adverse affect on any of the above safety related systems. Evaluations were performed to determine that all applicable regulatory requirements are satisfied. There is no effect on system and q
component functional design and safety bases as defined in the FSAR. There is no effect 4
on plant Emergency Operating Procedures (EOPs). Plant Technical Specifications have been I reviewed to assess the effects on applicable Limiting Conditions for Operation, Limiting l Safety System Settings, Safety Limits, and reactor thermal parameters and it has been concluded that the 85 F lake temperature does not reduce the margin of safety as defined ,
in the bases _for the Technical Specifications. No changes in the Technical Specifications are required due to operation with a lake water temperature of 85*F. The design basis lake i temperature limit will be raised from 82 F to 85 F.
Based on the preceding evaluations, it is determined that plant operation at current core l thermal power of 2536 MWT with an 85'F lake temperature does not involve an
. unreviewed safety question as defined in 10CFR50.59.
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ATTACHMENT I 69nual Summary of Channes. Tests, and Experiments for 1997/1998 JAF-SE-96-049, REV. 3: NEW AP-14.01, REVISION TO FSAR SECTIONS 9.8.5 AND 13.2.1 AND REVISION TO THE FIRE PROTECTION PLAN MODIFICATION: N/A The purpose of this safety evaluation is to support changes associated with a) Issuing a new Site Administrative Procedure AP-14.01, "JAF Fire Protection", b) Revising the Fire Protection Plan and defining those procedures / documents that implement and support the Fire Protection Program, c) revising the James A. FitzPatrick Final Safety Analysis Report (FSAR) to define the organization, responsibilities, and administrative controls which comprise the Fire Protection Program for the James A. FitzPatrick Nuclear Power Plant, and d) responsibilities within the Fire Protection Program The changes involve a) invoking administrative control and identifying the fundamental documents which support the Fire Protection Program, b) impose a safety review in accordance with MCM-4 if changes are made to those fundamental documents which constitute the Fire Protection Program and c) defining responsibilities for the Fire Protection Program.
The evaluation concluded that the proposed changes do not involve any unreviewed safety questions.-
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ATTACHMENT I Annt.al Summarv of Channes. Tests. and Experiments for 1997/1998 JAF-SE-96-052, REV. 2: CYCLE 13 RELOAD CORE AND CONTROL ROD BLADE REPLACEMENT MODIFICATION: M1-95-077 This evaluation documents the changes made to the reactor core by the refueling prior to the start of Cycle 13. The safety analyses performed on the Cycle 13 core have been reviewed and found adequate in scope and content. Fuel design changes have been i reviewed against designs previously used at FitzPatrick. On the basis of this review no l unreviewed safety question exists and the safety analyses properly support the limits l established in the Core Operating Limits Report (COLR).
Based on the foregoing summary, it is concluded that operation of FitzPatrick in the Cycle i
13 to the limits given in the COLR does not constitute an unreviewed safety question.
l The fifteen replacement control rod blades (CRB) installed under this modification have been evaluated against the SAR. There are no Technical Specification requirements
! affected by this modification. Design differences between these CRBs, other replacement CRBs and the original equipment were found to be insignificant to their interface with l other plant components and their function. Based on these conclusions the replacement CRBs are acceptable.
l l This revision of the safety evaluation supports the use of the revised Exclusion Region
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for full power operation for the remainder of the current reload cycle, in order to support l full power operation, plant procedures and other documentation must be revised to include the Exclusion Region consistent with Figure 8.4 of the COLR as approved in Revision 1 to this evaluation. A revised Buffer Zone is included where appropriate.
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1 ATTACHMENT I Annual Summarv of Chana-s, Tests, and F== iments for 1997/1998 JAF SE-96-067, REV.1:
REVISE INVENTORY AND EXTEND INSPECTION INTERVALS FOR EXTERIOR HOSE HOUSES AND CURB VALVES MODIFICATION: N/A This NSE justifies extending the frequency of exterior hose house inventory inspections and hydrant curb valve position inspections (limited to those located inside hose houses) from monthly to yearly, with an inspection of a controlled tamper seal on the hose house perfr,;med quarterly. In addition a quarterly inspection for physical damage and obstructions has been justified.
Thls NSE describes equipment required to be stored in exterior hose houses and defines the balance of equipment as optional. FSAR Section 9.8.3.7 will be revised to delete the sentence, "A sealed beam battery powered hand held light is provided at each exterior hose house." FSAR Section 9.16.3 will be revised to delete the sentence, " Sealed beam battery powered hand lights are provided at each exterior fire hose house."
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The revised exterior hose house inventory and inspection interval will ensure proper fire fighting equipment is available in the event of a fire and proper position of hydrant curb valves, and therefore there is no increase in the consequences of any fire or any other event. This equipment is not an initiator of any event nor does if affect any equipment I
important to safety. This change cannot increase the probability or consequences of any event or malfunction of any equipment important to safety. This change does not adversely sffect the ability to achieve and maintain safe shutdown in the event of a fire. Therefore, this safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10CFR50.59.
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ATTACHMENT I Annual Summa y of Chanaes. Tests, and Experiments for 1997/1993 JAF-SE-96-071, REV.1: IMPACT OF INCREASED ISOLATION TIME OF REACTOR BUILDING VENTILATION SYSTEM ON FSAR ANALYZED EVENTS MODIFICATION: N/A The purpose of this Nuclear Safety Evaluation is to review and analyze design basis function concerns identified with the secondary containment and the Reactor Building isolation and Control System (Refuel Floor exhaust radiation rnonitoring system) and determine if any unreviewed safety questions exist.
Operational restrictions have been placed on the Reactor Building ventilation system due to I a potential for unfiltered ground level release of radioactive material that may result from possible fuel damage and subsequent radiation releases during fuel handling activity periods.
The evaluation of this concern concludes that, for the small fraction of the Fuel Handing Accident source term to be released unfiltered at ground level before the completion of a ;
Reactor Building isolation, the dose to receptors in the low population zone and site i boundary are less than 25 percent of the 10 CFR 100 dose guidelines, and dose to receptors in the Control Room does cause an overexposure. The criteria for the Standard Review Plan,10 CFR 100, and General Design Criteria 19 are met. Therefore, an j unreviewed safety question does not exist. !
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ATTACHMENT 1 i Annual Summary of Channes Tests, and Experiments for 1997/1998 JAF-SE-96-073, REV. 0: UPDATE OF FSAR FIGURE 11.2-2 TO REFLECT 20TK-218 LEVEL SWITCH AS 20LS-357-1 AND 20LS-357-2 MODIFICATION: N/A The purpose of this Nuclear Safety Evaluation was to support changes to FSAR Figure 11.2-2 to correct inaccurately identified Radwaste Building Floor Drain Tank Level Switches 20LS-357-1 and 20LS-357-21 The update of the FSAR figure corrects the FSAR to reflect the two level switch configuration for the drain tanks level control system which was installed during plant construction. The function, operation, controls, or setpoints of the level control system are not changed.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10CFR50.59.
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l ATTACHMENT I Annual Summary of Channes, Tests, and Experiments for 1997/1998 JAF-SE-96-074, REV.1: OFF-GAS SYSTEM TROUBLESHOOTING MODIFICATION: N/A The purpose of this Nuclear Safety Evaluation is to evaluate proposed troubleshooting activities on the Off-Gas system during plant start up. This troubleshooting activity is l necessary to determine the position of the Off-Gas discharge valve (01-107AOV-100) and correct it if necessary.
This safety evaluation concluded that these activities, controlled by a temporary operating procedure, do not constitute an unreviewed safety question pursuant to 10CFR50.59.
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ATTACHMENT I l Annual Summary of Channes Tests, and Experiments for 1997/1998 l
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l JAF-SE-96-077, REV.1: SUPPLYING RECIRCULATION PUMP SEAL PURGE FLOW FROM AN ALTERNATE SOURCE !
i MODIFICATION: N/A r
The purpose of this nuclear safety evaluation was to determine the acceptability of providing a temporary supply of seal purge to the recirculation pumps for the purpose i of repairing / replacing a seal water flow control valves. The Recirculation pump seal water.does not perform a safety design function and is used primarily to extend seal 1 life. The accidents which assume Recirculation pump flow will not be impacted by this
activity. The worst possible consequence is a catastrophic failure of both seal stages which is bounded by the LOCA analyses. Therefore, providing a temporary supply of seal purge flow does not present an unreviewed safety question. 1 l'
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l ATTACHMENT I Annual Summary of Chances. Tests, and Experiments for 1997/1998 l
I JAF-SE-96-080, REV. 0: LONG TERM ISOLATION OF RADWASTE FLOOR DRAIN FILTER (20F-30)
MODIFICATION: N/A l The proposed activity is to isolate the Radwaste Floor Drain Filter for an extended period of time (years) due to failure of the filter septa. The design of the filter septa causes filter fouling after short periods of operation. The filter septa have failed due to high solids loading of the filter. Due to the poor design of the filter septa and lack of availability of a better design, the filter septa have not been replace'.
Isolation of the filter has no affect on evaluated accidents. There is no credit taken for the filter or the isolation valves in the existing accident analyses. 20F-30 performs no safety-related functions and the operation or lack of operation of this filter does not impact safety-related systems. The radwaste system is not safety-related except for the containment isolation valves and associated components which are not affected by the operation or lack i of operation of the filter. Isolation of the floor drain filter does not constitute an 3 unreviewed safety question.
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ATTACHMENT I Annual Summarv of Channes. Tests, and Experiments for 1997/1998 JAF-SE-97-001, REV. 0: RUNNING THREE TRAIN HYDROGEN INJECTION INTO TWO CONDENSATE BOOSTER PUMPS MODIFICATION: N/A The hydrogen injection system normally runs with three trains injecting into three condensate booster pump suction lines. To allow maintenance on the train check valves while maintaining hydrogen flow as high as possible, three trains willinstead be run into two booster pump suction lines. The train low hydrogen pressure trip will be defeated for the train under maintenance. The out of service train will also be purged with nitrogen prior to maintenance and after closing the system up. It will then be purged with hydrogen prior to re-commencing hydrogen injection with that train. The purge discharge will be via the installed hydrogen system roof vent.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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ATTACHMENT I Annual Summary of Channes. Tests. and Exaariments for 1997/1998 JAF-SE-97-002, REV. C: INSTALLATION OF AN UNDERWATER HIGH-FREQUENCY FISH DETERRENCE SYSTEM MODIFICATION: F1-96-030 The purpose of this NSE is to evaluate the interaction and safety aspects of an high-
- frequency fish deterrence system out on the plant's intake structure with the existing plant )
l configuration. This acoustic system is needed to maintain the SPDES permit with the j l D.E.C. commitment for May 2,1997. This modification attaches nine projector housings )
l and associated cable channel to the top of the intake structure. Existing cabling will be
!' reused to provide power from the screenhouse. The screenhouse will have a new panel i
to monitor the underwater equipment. '
This modification does not alter the design, operation, or function of the intake structure. '
! All of the interfaces with the intake structure will be designed as QA Category 11/111 with I l the understanding that its failure cannot impact operation of the intake tunnel.
l The modification does not change the design basis of the tunnel, intake deicing heaters,
- i. or circulating water system. No new obstruction have been created to effect any existing plant system.
l The installation of nine projector housings and associated cable channel on top of the intake structure with the use of spare cabling back to the screenhouse will not introduce j - new failure mechanisms as demonstrated in the review and analysis section of this NSE. l All of the cabling separation criteria will be maintained, therefore the modification cannot result in any new accident types. This modification does not interact with any safety
! system other than the attachment of the top of the intake structure. No new malfunction of a different type other than any evaluated previously in the FSAR can occur. All of the Technical Specification requirements for the existing bar heaters are maintained. The fish deterrence system could have no impact on the Technical Specifications. The physical l
changes made to the plant were evaluated and has been demonstrated that these changes will not have an adverse impact on the plant. This modification will not result in any l rebases to the environment and maintains the FSAR analysis. l I
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ATTACHMENT I Annual Summary of Channes. Tests. and Experiments for 1997/1998 L
JAF SE-97-003, REV. 4: USE OF REACTOR BUILDING CRANE TO MOVE IRRADIATED FUEL ASSEMBLIES IN THE SPENT FUEL POOL MODIFICATION: TEMPORARY MODIFICATION 97-056 l
This evaluation was prepared to demonstrate that irradiated fuel assemblies can be safely moved within the spent fuel pool using the Reactor Building Crane 1000 lb hoist to storage locations that are not accessible with the main hoist on the refuel bridga. Movement of the fuel assembly with the 1000 lb hoist will be performed using equipment and procedures similar in safety features to those used to move irradiated fuel with the main grapple on the refuel bridge. The evaluation addresses safety considerations involving release of radioactive material from damage to fuel during handling, inadve; tent exposure to personnel i l
by raising irradiated fuel out of the water or too close to the pool surface, and inadvertent criticality by placing the fuel in configurations not analyzed for criticality. This activity will ;
not be performed with the spent fuel pool gates open, and therefore the refueling interlocks ,
are not required on the 1000 lb hoist. Under these conditions this activity does not I constitute an unreviewed safety question based on the FitzPatrick Licensing Basis and the criteria in 10 CFR 50.59.
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ATTACHMENT I Annual Summarv of Channes. Tests, and Exneriments for 1997/1998 JAF-SE 97-OO4, REV.1: REVISION TO FSAR TO ELIMINATE THE RADWASTE BUILDING EXHAUST INLET ISOLATION DAMPERS MODIFICATION: N/A The purpose of this NSE is to address and evaluate a revision to FSAR Figure 9.9-4 and Section'9.9.3.5 to reflect current Radwaste Building Exhaust System configuration. Manual
. inlet isolation dampers are shown on FSAR Figure 9.9-4 but are not installed in the exhaust ductwork. This condition has existed since FSAR inception stating the dampers function is to provide filter isolation for maintenance with continued system operation. Revision 1 of this NSE provides clarification as to which filters are required to have the Radwaste Ventilation secured for filter maintenance.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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r ATTACHMENT 1
, Annual Summary of Chances. Tests, and Experiments for 1997/1998 i
l JAF-SE 97-005, REV. 0:
FEEDWATER FLOW ULTRASONIC MONITORING SYSTEM I (LEFM) !
MODIFICATION: M196-061 l The purpose of this Nuclear Safety Evaluation is to ensure that Modification M1-96-061, implementation of the Caldon Leading Edge Flowmeter (LEFM) component 06FM-100 does I not involve an unreviewed safety question.
The LEFM system will be used to establish a correction factor (CF) to the existing feedwater flow value input it.to 3D Monicore to compensate for the fouling bias at the feedwater venturis. This will allow the plant to operate closer to the licensed core thermal power limit. Feedwater flow indication input into 3D Monicore is higher than actual due to the fouling bias error. Compensating the feedwater flow input into 3D Monicore will be a manual process. The CF will be manually entered into EPIC which corrects the feedwater flow value input into 3D Monicore. Determination of the CF will be controlled by RAP i 7.3.37 and will consist of a process to validate LEFM flow values. Trending and analysis of I LEFM data will be performed to ensure LEFM system health.-
Installation of the LEFM system has no affect on any plant component or system. The feedwater flow CF values will not affect the feedwater control system. Installation will have no affect on any equipment important to safety identified in the SAR nor will it affect any accident analysis.
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ATTACHMENTI l Annual Summary of Channes. Tests. and Experiments for 1997/1998 l
JAF SE 97-006, REV. 0:
MANAGEMENT POSITION TITLE CHANGES: ELIMINATION OF POSITION OF PLANT MANAGER AND RE-ESTABLISHMENT OF I POSITIONS OF SITE EXECUTIVE OFFICER AND GENERAL MANGER - OPERATIONS MODIFICATION: N/A The organizational changes described in this NSE involve the elimination of the I position title of Plant Manager and re-establishing the positions of Site Executive Officer and General Manager - Operations with attendant minor revisions in
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reporting relationships. These changes reassign responsibilities but do not 1 eliminate any functional requirements. The changes are administrative in nature and do not involve plant equipment or operating conditions. The changes will (1) give the Site Executive Officer the direct authority over all activities at the site with the exception of Design Engineering, Quality Assurance and Information Technology and (2) provide an additional experienced manager focused on the integrated operating condition of the plant.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59. ;
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4 ATTACHMENT I Annual Summary of Channes, Tests, and Experiments for 1997/1998 JAF-SE-97-OO7, REV. 0: SERVICE PLATFORM REMOVAL MODIFICATION: M1-92-199 The purpose of this proposed activity is to permanently remove the service platform from the Reactor Building refuel floor and to update various sections of the FSAR to reflect the actual plant condition / configuration. The implementation of this activity will not cause any accidents or malfunction of equipment important to safety as described in the JAF FSAR. No new accidents or equipment malfunction scenarios could be created by this activity. In addition, the removal of the service platform does not increase the radiological consequences of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR nor can it cause any new consequences of an accident or malfunction of equipment important to safety. The proposed activity to remove the service platform from the Reactor Building does not adversely affect the JAF Technical Specifications and does not make the Technical Specifications inaccurate.
i in conclusion, the remov I of the service platform from the Reactor Building and the update of the FSAR to reflect the removal of the service platform is acceptable since this activity does not constitute an unreviewed safety quee u pursuant to 10CFR50.59.
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ATTACHMENT I Annual Summary of Chanaes. Tests, and Experiments for 1997/1998 JAF SE 97-008, REV. 0: JAF UPDATED FSAR SECTION 9.17 REVISION MODIFICATION: N/A The purpose of this Nuclear Safety Evaluation is to address and evaluate the changes to the JAF FSAR Section 9.17. These changes will either reflect the current configuration, or more accurately, reflect the manner in which refuel floor activities are performed to complete a refueling outage. Some of the changes are due to equipment changes while others are due to equipment retirement. Many of the changes are due to the use of upgraded equipment and other changes that have been implemented to reduce worker radiation dose.
The reactor servicing tools and equipment do not have any safety-related functions. The changes discussed do not have any impact on any other associated safety-related equipment or functions. The JAF FSAR section 9.17 descriptions of the reactor servicing tools do not describe any safety-related functions. The descriptions are intended to provide a description of how the refueling process can take place, not to restrict the means in which the process does takes place. The resulting changes to the FSAR from this safety evaluation will be written to more accurately describe the means of performing the affected reactor servicing tasks.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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~ 1 ATTACHMENT I Annual Summary of Channes. Tests, and Experiments for 1997/1998 l JAF-SE-97-OO9, REV. 0: MANAGEMENT POSITION TITLE CHANGES: QUALITY l
ASSURANCE DEPARTMENT ORGANIZATIONAL CHANGES MODIFICATION: N/A The organizational changes described in this NSE involve the elimination of the position of WPO-QA Manager, changes in QA organizational structures of both WPO-QA and IP-3 QA, the corporate officer position title change from Vice President-Management Information Services to Chief Information Officer, the corporate position title change from Manager -
Security to Director - Security, and the transfer of the SPEAKOUT Program responsibilities from WPO-QA Manager to Director - Security. These changes reassign responsibilities but do not eliminate any functional requirements. All administrative and operational functions previously performed continue to be performed. The Director - Security and the Director -
QA both report directly to the Vice President Appraisal and Compliance Services. The reorganization is designed to improve communication, responsiveness, and effectiveness of QA by consolidating functional lines of responsibility. The changes are administrative in '
nature and do not involve plant equipment or operating conditions. The changes do not affect the independence or scope of the QA organization or reduce QA Program commitments.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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ATTACHMENT I Annual Summary of Chances, Tests, and Experiments for 1997/1998
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l JAF-SE-97-010, REV. 0: TEMPORARY SURVEILLANCE TEST FOR 76P-4 MODIFICATION: N/A The purpose of this Nuclear Safety Evaluation is to evaluate testing of the East Fire pump 76P-4 through the test line for 76P-1 and 76P-2.
Original design of the test line for the East Diesel Fire Pump (76P-4) causes a significant amount of flow turbulence at the flow transmitter. This prevents repeatable readings for determining pump performance at all points on the pump curve. To provide a more accurate flow reading the test line for 76P-1 and 2 will be utilized. During the performance of this test the automatic start feature of 76P-1 and 2 as described in section 9.8.3 of the SAR will not be available. Recovery actions for 76P-1 and 2 will entail shutting the test valve 76 FPS-69 and restoring the switches for both pumps to the " automatic" position. During the performance of TST-72 an operator will be stationed at valve 76 FPS-69. Switch operation can be performed at the local fire panel.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 60.59.
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ATTACHMENT I Annual Summary of Chanaes. Tests, and Experiments for 1997/1998 JAF-SE 97-014, REV. 0: REPLACEMENT OF 02-2FCV-46A & 02-2FCV-46B MODIFICATION: M1-97-063 The purpose of this activity is to address tha replacement (installation and testing) of the existing adjustable flow control valves in the mini-purge seal lines to the Reactor Recirculation pumps (02-2P-1 A & 02-2P-18). The existing flow control valves are binding due to contamination resulting in a reduction in the seal water flow. The replacement flow control device will not have a manual flow adjustment. This modification removes the installed adjustable FCVs and replaces them with flow control devices with a constant flow rate. The new configuration includes inlet and outlet isolation valves as well as a bypass line and valve for manual flow control to allow for on-line cleaning and replacement of the FLOSERT. The purge seal line is designated QA Category ll/Ill. The failure of the mini-purge seal line is encompassed by the loss of recirculation pump transients and the LOCA. The removal of flow adjustment in the flow control valve will not change or affect the operation of the Reactor Water Recirculation or safety functions of the Control Rod Drive Hydraulic system. The changes proposed by this modification do not degrade the security plan, Quality Assurance Program, or the fire protection plan.
This change does not constitute an Unreviewed Safety Question per 10CFR50.59.
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ATTACHMENT I Annual Summary of Chanaes. Tests, and Experiments for 1997/1998 JAF-SE-97-015, REV. 0: REMOVAL OF ORIGINAL AIR DRYER AND OTHER INSTRUMENT AIR SYSTEM CHANGES TO THE FSAR MODIFICATION: N/A This Nuclear Safety Evaluation is performed to correct a number of differences between the FSAR and current plant conditions as reflected in JAF design documents for the Instrument Air System.
The update of SAR Section 9.11 and Fig. 9.11-1 is to correct design data and reflect current operation of the instrument air system These changes will not affect the operation of the two new 500 scfm air dryers each capable of providing full power plant instrument and breathing air requirements. The valving out of the original undersized plant 250 scfm air dryer will not adversely affect plant safety since the two new 500 SCFM air dryers are highly reliable and redundant.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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ATTACHMENT I Annual Summary of Channes Tests, and Experiments for 1997/1998 JAF-SE-97-016, REV. 0: REVIEW OF RAW WATER TREATMENT CLEARWELL LEVEL INSTRUMENT (42LI-108)
MODIFICATION: M1-91-025
' The clearwell level instrument (42LI-108) was removed before 1/9/91. The results of this safety evaluation have shown that the clearwell level instrument (42Ll-108) removal did not result in an unreviewed safety question. The effect of removing the clearwell level instrument (42Ll-108) on the water treatment system will be to remove a local level indicator function that is not required for safe or efficient operation of the water treatment system. The results of this safety evaluation has also shown that the clearwelllevel instrument (42Ll-108) removal did not result in a degradation in the security plan, OA program or the fire protection program. Therefore, the removal of the clearwell level instrument (42LI-108) is acceptable.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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L ATTACHMENT I i
Annual Summary of Channes. Tests. and Experiments for 1997/1998 l-JAF-SE-97-017, REV. 0: PROPOSED CHANGES TO SECTIONS 11.5.1 AND 11.5.2 OF THE FSAR MODIFICATION: N/A The purpose of this Nuclear Safety Evaluation is to show that an update to FSAR Section 11.5 does not constitute an unreviewed safety question.
The purpose of Section 11.5 of the Final Safety Analysis Report is to establish the radiation shielding design criteria and other plant features in order to 1) maintain the regulatory dose requirements for operating personnel and the general public during normal operating and A. emergency conditions; and 2) protect materials and components from radiation damage.
The proposed FSAR changes reflect current practice at the JAFNPP for design basis shielding and radiation protection of occupational personnel and the general public. These changes do not modify structures, systems, or components and do not cause an increase in radiation dose / dose rates.
Entrance to high radiation areas is controlled in accordance with regulatory and technical specification requirements. The revisions do not alter these requirements, but do define how access is administratively controlled.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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FE 1 ATTACHMENT I Annual Summary of Channes. Tests, and Experiments for 1997/1998
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JAF-SE-97-018, REV. 0: CRD FLANGE CAP SCREW SUPPLEMENTAL ACCEPTANCE CRITERIA MODIFICATION: N/A This Nuclear Safety Evaluation addresses a supplement to the allowable indication ,
acceptance criteria for CRD Cap Screws. The supplement to the criteria does not compromise the factors of safety required by Section XI of the 1980 ASME Boiler and Pressure Vessel Code, which is the ISI Code of Record for the second ISI inspection ,
l interval. The function of the CRD Cap Screws is unaffected by the application of )
supplemental examination methods and techniques and a supplement to the allowable indication acceptance criteria.
The JAF FSAR was reviewed and FSAR Section 3.5 discusses the failure of the CRD main flange cap screws. In the event of failure of these cap screws, the CRD housing i supports described in the FSAR Section 3.8 limit CRD flange travel to prevent any J significant nuclear transient. The postulated Control Rod Drop accident (FSAR Section 14.6) will not occur on failure of the CRD flange cap screws due to the proximity to CRD housing supports which willlimit downward motion of the CRD housing. A postulated loss of coolant accident which would result from a CRD flange cap screw failure is enveloped by the existing LOCA analysis in FSAR Section 14.6.
This supplement of the CRD Cap Screw allowable indication acceptance criteria will not affect the operation of the CRD system or mechanisms and will not result in an unreviewed safety question.
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ATTACHMENT I Annual Summary of Char . Tests. and Fraariments for 1997/1998 JAF-SE-97-019, REV. 0: -JAF UPDATED FSAR SECTION 3.9.3 REVISION _
MODIFICATION: N/A l
A safety evaluation was prepared to evaluate the acceptability of updating the FSAR by r6 moving references to an orifice flow switch (FS 54) and an associated alarm. The switch was part of the initial Standby Liquid Control system design and was subsequently deleted by GE Errata and Addendum #1 to GEK-16633 prior to the flow switch being installed in the plant. A reference to FS-54 and the associated alarm was inadvertently included in the SAR. The proposed change reflects the existing physical configuration of the Standby ,
Liquid Control (SLC) system. '
The proposed FSAR change does not result in a physical change to the plant. The evaluation concluded that the flow switch is not necessary to functions of SSCs important to safety. SLC system and equipment functions and methods of performing functions are not affected by the proposed change. Accidents, malfunctions, equipment failures and consequences were analyzed with no negative impacts identified. No changes to plant operating parameters or plant effluents are created by this change.
The safety evaluation concluded that the change does not constitute an unreviewed safety question and it is acceptable to update the FSAR by removing references to an orifice flow switch and associated alarm from the SLC system description.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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p ATTACHMENT I Annual Summary of Channes, Tests and Experiments for 1997/1998 1
[[::JAF-SE-97420|JAF-SE-97420]], REV. 0: REVISE 125 VOLT DC POWER SYSTEM, FSAR FIGURE 8.7-1, SHEETS 1 OF 5 AND SHEET 2 OF 5 MODIFICATION: N/A l This safety evaluation was prepared to evaluate the acceptability of updating the FSAR to I make the 125 voit D-C Power System figures located in FSAR Section 8.7 consistent with JAF electrical drawings.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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ATTACHMENT I Annual Summary of Channes. Tests, and Experim6nts for 1997/1998 i
JAF-SE-97-021, REV.1: INSTALLATION OF ONE MAXIMUM DENSITY SPENT FUEL RACK IN THE POOL MODIFICATION: F1-96-008 This modification includes the installation of one (1) max-density rack in the Northeast corner of the Spent Fuel Pool (SFP). The cell locations in this rack (Rack N3), in addition to the utilization of the peripherallocations of the existing racks that are inaccessible with the J refuel bridge (113 locations) shall provide enough number of locations to accommodate the I discharge from RO13.
Racks N1 and N2, to be added after the SFP cleanup in 1999 and prior to RO14, will provide enough locations for the discharge from that outage.
If the dry cask storage facility (ISFSI) is not in place by 2001 four smaller racks designated as F1 through F4 will be fabricated and installed to accommodate the discharge from RO15. Thus, provisions are made for continued operation of JAF until 2004. Note that the full-core off-load capability is always maintained.
The installation of one or more of the seven racks in accordance with the plant schedule between 1998 and 2002, does not possess any concerns for the safe operation of the plant and does not pose any threat to the health and safety of the general public.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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p a ATTACHMENT I Annual Summarv of Cham. Tests, and F=-iments for 1997/1998 i
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[ JAF-SE-97-024, REV. 0: PROPOSED CHANGES TO THE FSAR SECTION 7.13.4, t
CONSTANT AIR MONITORS l
L MODIFICATION: N/A The purpose of section 7.13.4 of the Final Safety Analysis Report is to establish the basic function of portable continuous air monitoring. Portable continuous air monitor alarms provide an early warning to workers that airborne concentrations exist and/or have changed and they need to immediately leave the area to minimize potential intake.
The purpose'of this nuclear safety evaluation is to show that an update to FSAR Section 7.13.4 does not constitute an unreviewed safety question. The proposed revisions reflect current portable constant air monitoring technology and administrative control. The propesed revision is consistent with regulatory requirements, technical specifications, and !
procedural requirements at the James A. FitzPatrick Nuclear Power Plant. !
l l The detailed equipment specifications were removed from the FSAR and the purpose of i portable continuous air monitoring was added. This revision will allow for technological l advances while maintaining the basic functions for monitoring airborne radioactivity. INPO 91014, Regulatory Guide 8.25, and NUREG 1400 give guidance on when and how to provide monitoring, the types of air sampling to use and where to place air sampling equipment. JAFNPP procedures follow these guidelines.
j' Although the portable CAMS can be used with the effluent monitoring systems it is not the intent to replace safety-related effluent monitoring systems. The portable CAMS are used as early warning devices in the restricted area for the occupational workers.
The nuclide for calibrating the IM 1 A portable iodine CAM is Ba 133 and not 1-131. The CAM is calibrated to the Ba-133 356 kev peak and then adjusted to detect the 1-131 364 ;
kev peak.1-131 is not a practical calibration source due to its short half life therefore, Ba- '
133 is used for calibration. Ba 133 is recommended as the calibration source by the vendor technical manual.
The particulate CAMS can be used to monitor for fission products or activation products depending on where they are located in the plant. The particulate CAMS are calibrated to respond appropriately to either source term. .
l- The changes do not modify structures, systems, or components and do not cause an increase in radiation dose or dose rate. Therefore, these changes do not increase the probability of occurrence or the consequences of an accident or malfunction of safety related structures, systems or components previously evaluated in the FSAR. The proposed '
changes do not create the possibility of an accident or malfunction of safety-related structures, systems, or components of a different type than evaluated previously in the ,
FSAR. '
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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ATTACHMENT I Annual Surnmarv of Chances. Tests, and Experiments for 1997/1998 l
l JAF-SE-97-026, REV. 0: REMOVAL FO THE INTERNALS OF VALVE 12FCV 20 MODIFICATION: M1-97-023 1
The intervals of manual control valve 12FCV-20 were removed in accordance with Temporary Modification 96-058 and permanently removed by Modification M1-97-023.
The purpose of this safety evaluation is to evaluate the pcrmanent removal of valve internals and reflect the FSAR change associated with this change.
This modification meets the design code requirements and does not alter the function of the valve to maintain the pressure integrity of seal purge line as required per modification j F1-90-202. The modification does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report, nor willit create the possibility of an accident or malfunction of a different type than previously evaluated in the safety evaluation report.
The margin of safety as defined in the basis for any technical specification is not reduced.
l The modification does not constitute an Unreviewed Safety Question (USQ). The modification activities will have no impact on any equipment important to safety.
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ATTACHMENT I Annual Summarv of Chances. Tests, and Ex_aariments for 1997/1998 JAF-SE 97-027, REV. 0:. PRESSURE TEST OF CORE SPRAY SUCTION PIPING FROM
~ CONDENSATE STORAGE TANKS (CSTs) j MODIFICATION: N/A
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The purpose of this Nuclear Safety Evaluation (NSE) is to evaluate the pressure test of core spray suction piping from CSTs per TST-74. During the test, a temporary test connection will be installed on the drain pipe from 14 CSP-735A per WR 96-04200-02. The original piping will be restored at the conclusion of TST-74. However, the drain pipe from 14 CSP- l 735A will be temporarily altered and.the core spray suction piping from CSTs will be temporarily out of service. Therefore, this safety evaluation is needed. The test procedure consists of the following major steps:
The proposed activity, pressure testing of the core spray pump suction lines from CSTs, will only test the piping between the safety-related suction line isolation valves,14 CSP-8A, 14 CSP-8B, and the QA category ll/Ill CSTs Valves 14 CSP-8A and 14 CSP-8B will be locked closed to ensure the operability of the safety-related core spray system injection lines.
The suction line will be drained using small bore valves 14 CSP-735A and 14 CSP-735B before opening 14 CSP-15 to prevent the inadvertent draining of the CSTs to the crescents.
Valves 14 CSP 1 A,14 CSP-1B,14 CSP-8A and 14 CSP-8B will be verified to be reasonably leak tight before proceeding with the TST-74. This will prevent excessive boundary leakage during the test.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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ATTACHMENT l '
Annual Summarv of Channes. Tests. and Experiments for 1997/1998 i
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i JAF SE 97-028, REV. 0:
' DOCUMENT ALPS DEMINERALIZER SYSTEM (MFTDS) AS PERMANENT INSTALLED NYPA EQUIPMENT l MODIFICATION: M197-038 This Nuclear Safety Evaluation documents that there is no Unreviewed Safety Question
! associated with this modification which formally declares and re classifies the MFTDS as permanent NYPA installed equipment. It does not degrade the Safety Design Bases of the l Liquid Radwaste System (20) in any way that is stated in FSAR Section 11.2 nor does it l impact in any way the previous safety evaluation, reference 7.3. This modification does not :
result in a change to the Technical Specifications nor does it negatively impact in any way I other environmentally qualified systems / structures or Fire Protection Systems.
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ATTACHMENT I Annual Summary of Cha.tmes. Tests, and Experiments for 1997/1998 i JAF SE-97-029, REV.1: ON-LINE CLEANING OF THE CONDENSATE STORAGE TANKS MODIFICATION: N/A Removal of a CST from service for maintenance during unit operation is consistent with the plant design and UFSAR. CST cleaning can be accomplished without impairing HPCI or RCIC operability per the plant Technical Specifications and is consistent with the various design documents. Adequate inventory will be maintained in the in-service CST to ensure compliance with the SBO rule during cleaning operations. Appropriate FME controls will be utilized throughout the cleaning and inspection operation.
The radiological consequences of postulated liquid leakage from the Chem-Nuclear cleaning apparatus is bounded by the existing UFSAR analysis. The valve manifolds and pumps used during cleaning operations will be in containers to reduce the potential for release of tank fluids or the resin /studge resident in the tanks.
Based on the above, cleaning of the tanks is acceptable, does not represent an unreviewed safety question, and is consistent with the plant design.
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ATTACHMENT I Annual Summary of Channes. Tests, and Experiments for 1997/1998 JAF-SE 97-030, REV. 0: REVISION TO AP-14.01, FIRE PROTECTION PROGRAM *, AND JAF-RPT-FPS-02367, FIRE PROTECTION PLAN MODIFICATION: N/A The proposed changes are to revise JAF Administrative Procedure AP-14.01, "JAF Fire Protection" to identify the procedure used to perform Fire Protection / Appendix R compliance reviews and JAF-RPT-FPS-02367, "JAF Fire Protection Plan", to reflect that a 3 fire protection implementing procedure (FPP) has become an Administrative Procedure l
(AP), that several FPPs have been consolidated, that several FPPs outside the auspices I of the NRC fire protection program have been removed and that several new procedures have bean added, either to enhance the program or to provide clarity.
These procedural changes cannot increase the probability of or consequences of any event or malfunction of equipment important to safety. This change does not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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ATTACHMENT I A_qnual Summary of Chances. Tests, and Experiments for 1997/1998 JAF-SE 97-031, REV. 0: PIPE STRESS ACCEPTANCE CRITERIA FOR THE ANALYSIS OF i ISOLATED WATER-FILLED PlPlNG SECTIONS SUBJECTED TO THERMAL PRESSURIZATION DURING FAULTED CONDITIONS
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MODIFICATION: N/A The purpose of this evaluation is to demonstrate that the use of plant specific acceptance criteria for qualifying isolated water filled piping sections subjected to thermal pressurization during a Design Basis Accident (DBA) condition is acceptable.
These plant specific requirements are necessary since neither ANSI B31.1 Power Piping Code 1967 Edition through 1969 Addenda the JAF code of record, nor the "JAFNPP Design Criteria for Balance of Plant (BOP) Piping Stress and Supports" addresses i qualification of piping subject to tb.ermal pressurization.
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This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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ATTACHMENT I Annual Summary of Chaneee Tests, and F Mments for 1997/1333 JAF-SE-97-032, REV. 0: INSTALLATION OF RHR SHUTDOWN COOLING SUCTION HIGH POINT VENT LINE MODIFICATION: M197-021 This Nuclear Safety Evaluation documents that there is no Unreviewed Safety Question i.
associated with this modification. Minor Modification Mi-97-021, which is being generated in response to LER 97-002, Rev. O, will permanently install a 3/4" RHR SDC high point vent l line on RHR pump suction piping. This high point vent line will improve the system venting l capability and will be opened to support plant operations and maintenance. This valve will i be closed during normal RHR system operation. This modification does not degrade the design bases of the Residual Heat Removal System in any way that is stated in both the l {
JAF Design Basis Document ([[::JAF-DBD-010|JAF-DBD-010]], Rev. 0) and the FSAR section 4.8. This
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modification does not result in a change to the Technical Specifications nor does it l
negatively impact in any way other environmentally qualified system or structures. ]
y This safety evaluation concluded that this activity does not constitute an unreviewed
. safety question pursuant to 10 CFR 50.59.
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I ATTACHMENT I Annual Summary of Channes, Tests, and Experiments for 1997/1998 !
JAF-SE-97-033, REV. 0: RCIC LOGIC BUS A POWER MONITOR TEST SWITCH MODIFICATION: M195-096 The purpose of this Nuclear Safety Evaluation is to support changes as identified in modification M195-096.
The purpose of this activity is to install a manual,2 position, key-lock test switch for testing of the RCIC Logic Bus A Power Monitor. This switch will replace the removal and re-insertion of fuse 13A-F34 currently required under surveillance test ST-2M (ECCS Trip Systems Bus Power Monitors Functional Test). The installation and testing of the loss of power annunciator does not effect the operation of the RCIC system. This installation will not effect the operation of the plant or its systems in any way not previously evaluated in the SAR. i This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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ATTACHMENT I
' Annual Summary of Chanaes, Tests, and Experiments for 1997/1998 l l
j JAF-SE-97-035, REV. 0: OPERATION OF MOTOR GENERATOR VENTILATION SYSTEM WITH LESS THAN 6 GRAVITY ROOF VENTILATORS MODIFICATION: N/A l
The purpose of this nuclear safety evaluation was to determine the acceptability of allowing maintenance to be performed during periods of cool weather. This will remove the reference to "6" gravity dampers in section 9.9.3.3 of the FSAR. Procedures will be in place to monitor the area temperatures and take actions to preclude the temperature going above the design range. Therefore, allowing the system to be shutdown for maintenance does not present an unreviewed safety question.
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m ATTACHMENT I Annual Summary of Channes. Tests, and ExnMments for 1997/1998 JAF-SE 97-036, REV. 0: FILTERING EMERGENCY DIESEL GENERATOR (EDG)
UNDERGROUND FUEL OIL STORAGE TANK FUEL MODIFICATION: N/A The purpose of this Nuclear Safety Evaluation is to review and assure these proposed activities do not involve an unreviewed safety question.
Filtering the contents of an EDG underground fuel oil storage tank during plant operation '
does not render the EDG inoperable. Adequate inventory will be maintained in the underground tank throughout the filtering activity. In the event of vendor equipment failure or severe weather (heavy rain, lightning , high winds, etc.) the filtering activity will be i terminated until conditions permit restart of the activity. The filtering equipment will be staged in the diked area designed to accept a fuel oil spill to minimize the consequences of a spill. In the event of an EDG demand, filtering activities will be halted and the system will !
be restored to normcl.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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4 ATTACHMENT I Annual Summarv of Channes, Tests, and Experiments for 1997/1998 JAF-SE 97-039, REV.1: TORUS /DRYWELL VACUUM BREAKER ALTERNATE TEST METHOD MODIFICATION: N/A The purpose of this Nuclear Safety Evaluation is to review and analyze the acceptability of the proposed changes in test methods to torus /drywell vacuum breakers.
Periodic testing of the pressure suppression chamber (torus)/drywell vacuum breakers requires equalizing the (intentional) differential pressure between the two containment air spaces. This evaluation supports a change to the method of accomplishing this test recondition. The existing UFSAR test description permits the pressure suppression function I to be bypassed prior to performing the required testing. The LOCA type of accident relies on the proper performance of containment air space energy paths to ensure acceptable i accident response. The proposed test method will improve containment loading in the event of a LOCA over the current test methodology while the equalization is in progress.
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The UFSAR description is being revised to support the preferred approach. The proposed change to UFSAR 5.2.3.6 reads: "(1) equalizing the pressure between the suppression chamber and drywell by using venting or make-up in any combination which does not create a continuous path between the pressure suppression pool and drywell air spaces".
The use of nitrogen make-up to the torus is supported up to the normal operating conditions documented in the FitzPatrick Containment Data Specification.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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ATTACHMENT I i Annual Summarv of Channes. Tests, and Experiments for 1997/1998 JAF-SE-97-040, REV. 0:
REORGANIZATION OF EMERGENCY RESPONSE SUPPORT AND RECOVERY RESPONSIBILITIES l
MODIFICATION: N/A 1
The purpose of this Nuclear Safety Evaluation is to review and assure these organizational l changes do not involve an unreviewed safety question, j I
The Headquarters (White Plains Office) Emergency Response / Recovery Organization was eliminated, including the associated Headquarters Emergency Response Support and Recovery Plan and implementing procedures and the Headquarters Emergency Response Center. It's functions have been distributed over a de-centralized Recovery Organization (RO) consisting of plant staff and additional support personnel. The changes do not eliminate any functional requirements or reduce the expertisc of the RO. The changes reflect the realignment of headquarters functions to the plant sites that has taken place during the past several years and will enhance the utilization of these site personnel for the support of emergency recovery activities. Existing corporate staff will be made available on an as-needed basis and will aport to an appropriate work location.
Changes in emergency response support and recovery organizations do not change the plant physically, and do not affect how the plant is operated, tested or inspected.
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ATTACHMENTI Annual Summary of Channes. Tests, and Experiments for 1997/1998 JAF SE-97-041, REV. 0: INSTALLATION OF A BONNET PLUG IN 31SV-100 MODIFICATION: TEMPORARY MODIFICATION 97-118 A safety evaluation was prepared to evaluate the occeptability of plugging a relief valve 1 bonnet vent to remove an air inleakage path to the condenser. The associated reduction in lift pressure was assessed against FSAR requirements and operating margins to gauge the l consequences. !
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The reboiler is not a factor in evaluated accidents. The effects of the proposed activity will affect only operating margins. This change will not increase the probability of occurrence or 1 consequence of an accident evaluated in the safety analysis report. No effects on equipment important to safety exist. i The system cannot create a new accident situation by the actions contained in this review.
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No new malfunctions are created as the system is being used in ways that reflect original j design considerations for loading and operation.
3 These system functions are absent from Technical Specification safety margins. No reductions of margin exist for this activity. There are no unreviewed safety questions or adverse effects on Fire, Security, Quality Assurance or Environmental Programs. !
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E ATTACHMENT I Annual Summary of Chanaes. Tests, and Experiments for 1997/1998 l
JAF SE-98-001, REV. 0: OPERATION OF REBOILER USING 31MOV-151
! MODIFICATION: N/A I
A safety evaluation was prepared to document the existing design feature of a bypass around the normal pressure control interface between the reboiler and the main turbine seal steam header. The review cycle identified an earlier error had been introduced in the turbine gland sealing system description. This section was restored to accuracy at the originallevel of detail.
The reboiler is not a factor in evaluated accidents. The effects of the proposed activity will !
affect only operating lineup options. This change will not increase the probability of l occurrence or consequence of an accident evaluated in the safety analysis report. No I effects on equipment important to safety exist.
The system cannot create a new accident situation by the actions contained in this l
review. No new malfunctions are created as the system is being used in ways that reflect original design considerations for loading and operation.
These system functions are absent from Technical Specification safety margins. No reductions of margin exist for this activity. There are no unreviewed safety questions or 1 adverse effects on Fire, Security, Quality Assurance, Environmental Programs, or the l Emergency Plan.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59, i
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l ATTACHMENT I Annual Summary of Channes, Tests, and Experiments for 1997/1998 JAF SE-98-002, REV. 0: SURVEILLANCE TEST (ST) REDUCTION PROJECT MODIFICATION: N/A This NSE justifies the extension of the frequency of the testing / inspection of specific fire protection systems / equipment. The extensions are based on a review of plant-specific past !
performance data which is contained in Evaluation No. JAF-RPT-FPS-02708. l This NSE demonstrates that extending the frequency of testing / inspection of fire protection systems / equipment will not adversely impact the reliability or availsbility of the systems / equipment. The technical basis and evaluation is contained in an engineering evaluation, JAF-RPT FPS-02708.
This activity does not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question purauaivt to 10 CFR 50.59.
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l ATTACHMENT I Annual Summary of Channes, Tests, and Experiments for 1997/1998 JAF SE 98-003, REV. 0: HANDLING AND STORAGE OF FUEL PIN WB00320 MODIFICATION: N/A The handling and storage of fuel assemblies (with the exception of casks) is a normal activity performed during all modes of operation. These evolutions are routinely performed using plant equipment designed for this purpose.
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The handling and storage of individual fuel pins, sections thereof, including fuel pellets, I while not a routine event, can be expected during the course of power plant operations.
This safety evaluation has demonstrated thatthe handling of fuel pin WB00320 does not ,
l constitute an unreviewed safety question. This conclusion is reached primarily through the I evaluation of the design basis fuel handling accident. This design basis event was l
compared to the possible consequences resulting from the activity of handling and storage l l
of fuel pin WB00320 and the design basis event was found to be bounding in all cases.
Since the design basis event is also a safety design basis for related plant systems which ensure that the limits cf 10 CFR 100 are not exceeded, the proposed activity (s) may be performed safely.
Other considerstions were given to maintaining spent fuel pool less than 0.95 K.,, during the proposed activity (s). This requirement, (Technical Specification 5.5. B), continues to be met. The personnel safety and radiological controls will be met through normally established procedural controls.
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l ATTACHMENT I Annual Summarv of Channes. Tests, and Exoeriments for 1997/1998 i
JAF SE-98-004, REV. 0: REPLACEMENT OF PCIS DISPLAY ON 09-4. PANEL ,
I l MODIFICATION: M197-051 Modification Mi-97-051 installs a new PCIS Display consisting of 104 General Electric ET-16 indicating lights mounted to a plate containing a separation barrier in 09-4 Panel.
Currently installed on the PCIS Display is a Staco switch matrix with Ledtronics LED indicating lights implementation will consist of the fabrication of a new display mounting plate utilizing ET-16 indicating lights and a ceparation barrier. To provide adequate spacing between lights the existing cut out in 09-4 Control Room Panel will be enlarged. The
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i mounting plate will be installed in the enlarged cut out and new SIS wire will be installed.
The PCIS Display provides direct open/ closed indication status of the Primary Containment Isolation Valves and is consistent with guidance of Regulatory Guide 1.97. The replacement of the existing LED lights with GE ET-16 incandescent provides more reliable indication of PCIS valve status. All previously evaluated DBA's in Chapter 14 and the applicable safety evaluations in the SAR have been reviewed and are not affected.
The purpose of this safety evaluation is to demonstrate that installing a different type of indicating light in the 52 valve control circuits does not introduce any new or unanalyzed faults or conditions. The GE ET-16 light meets the design, material and construction standards for QA Category I installation. System performance is unchanged, but indicating light reliability will be improved. Radiological consequences are not altered with the new i
type of light. All the PCIS sample valves are designed failsafe to maintain Containment isolation.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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1 ATTACHMENT 1 Annual Summary of Channes, Tests, and Exneriments for 1997/1998 I
JAF-SE-98-005, REV. 0: INSTALLATION OF A TEMPORARY SUMP PUMP IN THE REACTOR BUILDING EQUlPMENT DRAIN SUMP FOR MAINTENANCE PURPOSES MODIFICATION: TEMPORARY MODIFICATION 98-005 A safety evaluation was prepared to evaluate the acceptability of installing a temporary sump pump in the reactor building equipment drain sump while the permanent pump and associated instrumentation are removed to allow performance of maintenance activities.
The equipment drain sump is not a factor in evaluated accidents. The sump pump and associated instrumentation is not safety-related. The floor drain sump and pump located in the same room are a factor and will be undisturbed by the proposed activity No new accident scenarios will be created by performing the proposed activity. The temporary pump will take a suction from the same location and discharge into the same pipe as the normally installed pump. Power will be provided to the temporary equipment by a normal 120 volt AC outlet in the area of the sump.
The permanent sump pump is provided with automatic start and stop capability and a high level alarm. The temporary equipment will require manual operation and will net be equipped with an alarm. This is acceptable because in the event the sump overflows the water will flow to the floor drain sump in the area and be removed by that pump prior to affecting any other equipment in the crescent. An overflow of the equipment drain sump would not go undetected. Operations would notice the higher than expected pump out frequency of the floor drain sump or observe the overflow on operator rounds. There are alarms for floor drain sump pump out time and sump fill time.
The reactor building equipment drain sump equipment functions are not mentioned in any Technical Specification safety margins. No reductions of margin of safety exist for the proposed activity. There are no unreviewed safety questions or adverse effects on Fire, Security, Quality Assurance, Environmental Programs, or the Emergency Plan.
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ATTACHMENT I Annual Summarv of Channes, Tests, and Experiments for 1997/1998 JAF-SE-98-006, REV. 0: CYCLE 14 RELOAD CORE MODIFICATION: M197-030 This evaluation documents the changes made to the reactor core by the refueling prior to the start of Cycle 14. The safety analyses performed on the Cycle 14 core have been reviewed and found adequate in scope and content against the requirements in General Electric Standard Application for Reactor Fuel (GESTAR) NEDE-24011-P-A-13, August 1996 and the FitzPatrick SAR and Technical Specifications. Fuel design changes have been reviewed against designs previously used at FitzPatrick. On the basis of this review, no unreviewed safety question exists and the safety analyses properly support the limits established in the Core Operating Limits Report (COLR).
Based on the foregoing summary it can be concluded that operation of FitzPatrick in the Cycle 14 core as given in the reference documents to the limits given in the COLR does not constitute an unreviewed safety question.
The fourteen replacement control rod blades (CRB) installed under this modification have been evaluated against the SAR. There are no Technical Specification requirements affected by this modification. Design differences between these CRBs, other replacement CRBs and the original equipment were found to be insignificant to their interface with other plant components and their function. Based on these conclusions the replacement CRBs are acceptable.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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ATTACHMENT I Annual Summary of Channes, Tests, and Experiments for 1997/1998
- JAF-SE 98-009, REV. 0: PROVIDE A MEANS TO RELEASE THE PRESSURE OF THE DRYWELL DAMPER ACTUATORS MODIFICATION: M1-97-035 The purpose of this Nuclear Safety Evaluation is to demonstrate that the proposed activity does not involve any unreviewed safety questions.
The purpose of this activity is to reduce the excessive closing time of the Drywell cooler Dampers. The closing time will be significantly improved through the relocation and replacement of the existing pressure regulators. The new regulators will be OA Category I and Seismic. This modification will not affect the N2 pressure boundary established per Modification M1-93-020 Rev.1. The changes in this modification do not degrade the security plan, Quality Assurance Program, or the Fire Protection Plan. This change does not constitute an Unreviewed Safety Question per 10CFR50.59.
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ATTACHMENT I Annual Summary of Chanaes Tests, and Exneriments for 1997/1998 JAF SE 98-010, REV. 0: CARBON DIOX1DE PRESSURIZATION OF THE RELAY ROOM MODIFICATION: N/A This evaluation addresses the change in Reluy Room pressure and carbon dioxide concentration due to timer setpoint repeatabi!ity being different than that originally assumed.
This evaluation assessed the required carbon dioxide concentration in the relay room and the assurance that the carbon dioxide leakage into the Control Room is within acceptable limits. Calculations that use data collected during fan pressure testing were used to assess leakage through the Control Room electrical penetrations and structural components.
Allowing the motor operated damper to remain open for a longer period before closing following the discharge of the carbon dioxide system will not increase the probability of occurrences or consequence of an accident or malfunction of structures, systems, of components important to safety previously evaluated in the UFSAR. In addition, this change will not create an accident or malfunction of a different type than evaluated in the UFSAR. The Relay Room carbon dioxide fire suppression system will meet or exceed its design concentration requirements per NFPA 12 by alternate testing methodologies in lieu of a full discharge test. The system concentration has been previously demonstrated by tracer gas testing (Special Test Procedure STP-76AU). The system concentration has been analyzed with a calculation (Reference 26) to determine an acceptable open damper time.
Control Room habitability has been established based on a carbon dioxide short-term i
exposure limit of 3% by volume in the Control Room. A short-term exposure is considered to be about 15 minutes (Reference 28). The 3% by volume is acceptable for short-term exposure according to NFPA 12. NFPA 12, section 12-50 states that 3% or 4% carbon l dioxide level will cause rapid breathing, but will otherwise have no important effect on people for relatively short exposures.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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ATTACHMENT I i Annual Summary of Channes. Tests, and Experiments for 1997/1998 JAF-SE-98-011, REV. 0: DRYWELL CONTINUOUS AIRBORNE RADIOACTIVITY MONITORS (DW CAMS) EPIC ALARMS UPGRADE AND MISCELLANEOUS INTERNAL WIRING CHANGES j MODIFICATION: M197-048 The primary purpose of Modification M197-048 "Drywell Continuous Airborne Radioactivity Monitors (DW CAMS) EPIC Alarms Upgrade & Misc. Intemal Wiring Changes" is to upgrade the existing EPIC alarm circuits to include high radioactivity alarm capability in the Control Room for the Drywell Continuous Airborne Radioactivity Monitors (DW CAMS).
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- Modification M1-97-048 also resolves FSAR description discrepancies for alarm and recording capabilities in the Control Room and supports the draft implementation of the Improved Technical Specifications.
In addition, Modification M1-97-048 modifies the sample flow controllers' power source and output circuits. These changes have been included in the scope of this modification to resolve internal wiring inconsistencies between the two panels. The operational / functional aspect of these changes have been previously evaluated under Modification M1-89-072 " Misc. Design improvements to DW CAM Panels 04-1 and 04-2" and NSE No. JAF-SE-89-139.
The existing DW CAMS EPIC alarms (D357/D358) are being upgraded to provide high radioactivity and instrument failure alarms in the Control Room. The particulate (17RM-102A/8), iodine (17RM-101 A/B) and gaseous (17RM-103A/8) ratemeter existing alarm relay's spare contacts will be utilized for the new EPIC alarm circuit design. The high radioactivity alarm relay contacts and the instrument failure alarm relay contacts for each ratemeter channel will be wired in parallel to the existing flow disturbance (17FS-OO1 A/
B) alarm circuit branches. The parallel branches, tied together at the DW CAM Panels, will utilize the existing field cables to provide input to the EPIC Computer. The EPIC digital input j point identification numbers (D357LD358) will remain the same, however, their name, i description, reset and set tags will be updated accordingly to reflect the new alarm circuit design.
This safety evaluation has reviewed the proposed activities and concluded that these do not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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1 ATTACHMENT 1 l Annual Summary of Channes. Tests, and Exceriments for 1997/1998 JAF SE-98-012, REV. 0: MAIN STACK HEPA DRAIN MODIFICATION MODIFICATION: M1-97-078 Minor Modification M1-97-078, " Main Stack HEPA Filter Drain" will install a normally closed, solenoid operated drain valve (01-107SOV-700) with a remote push button switch (01-107PB-1) in a drain line including a loop seal (01-107-3/4"-DRW-152-107). The existing High Efficiency Particulate Air (HEPA) filter outlet piping provides no drain path. As such, during certain operating and environmental conditions, Off-Gas and Main Stack condensation can fill the outlet piping resulting in high differential pressure across the HEPA filter or fluctuating outlet pressures indicating a clogged or blocked condition. The addition of a drain line will allow the Off-Gas System to collect, process, hold and control the release to the atmosphere of the gaseous radioactive waste from the main condenser i air ejector system line. The Off Gas HEPA filter drain line is non-safety related and )
does not perform a function important to nuclear safety. The drain line is designated QA Category ll/ill. The changes proposed by this modification do not degrade the Security Plan, Quality-Assurance Program, the Fire Protection Program, the Environmental Report or the Emergency Plan. This safety evaluation has reviewed the proposed activity and concluded that this does not constitute an unreviewed safety question per 10CFR50.59.
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ATTACHMENT I Annual Summarv of Chanaes Tests, and Experiments for 1997/1998 JAF-SE-98-013, REV. 2:
RESIDUAL HEAT REMOVAL AND CORE SPRAY SUPPRESSION POOL SUCTION STRAINER REPLACEMENT MODIFICATION: F1-97-031
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Modification F197-031 Rev. O, " Residual Heat Remo/al and Core Spray Suppression Pool Suction Strainer Replacement", installs new suppressico pool suction strainers for Residual Heat Removal and Core Spray system pumps. The replacement strainers are substantially larger than the current strainers and provide additional margin for debris loading following a design basis accident. Consistent with the original licensing br. sis for FitzPatrick, the replacement strainerr ' ave been evaluated to ensure adequate NPSH would be available to !
the pumps, under the most limiting conditions analyzed in the current LOCA analysis, with ;
the suction strainers 50% blocked. The strainer assemblies and supports have been i analyzed and qualified for seismic and Mark I hydrodynamic loads specified in the FSAR l and Plant Unique Analysis Report.
A temporary construction opening will be made in the torus to facilitate personnel and !
equiprnent movement for the strainer installation. Integrity of the repaired opening will be I ensured by qualification and non-destructive examination of the weld, and by post modification testing. A primary containment pressurization test will be performed at peak accident pressure to verify leak-tightness at the weld and structural integrity of the repaired area.
A review of the modification in accordance with 10CFR50.59 concluded that the modification does not increase the probability or consequences of an accident or of a j malfunction of equipment important to safety previously evaluated in the safety analysis report. Further, the possibility of an accident or malfunction of a different type than any evaluated previously in the safety analysis report will not be created. The margin of safety as defined in the Technical Specification Bases is not reduced and no Technical Specification change is required. This change, therefore, does not involve an unreviewed safety question.
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i ATTACHMENT I .
Annual Summary of Chanaes. Tests, and Experiments for 1997/1998 '
JAF-SE 98-015, REV. 0: INSTALLATION OF HPCl/RCIC PACKAGE BOILER AND ASSOCIATED PIPlNG MODIFICATION: F1-97-034 Modification F1-97-034 involves the use of a temporary boiler and the installation of piping, pipe supports and valves to establish a steam supply flow path for HPCI and RCIC turbine slow roll testing and overspeed testing. The new piping associated with this modification goes through the Nitrogen Storage Building and into the Reactor Building through penetration S-733. It ties in to the existing HPCl and RCIC steam supply lines downstream of the outboard containment isolation valves, and uses a spectacle blind flange / spacer ring arrangement to establish and isolate the flow path. The piping from the steam supply tie-in to the upstream flange adjacent to the spectacle flange is designed to the same classification as the respective HPCI and RCIC steam supply lines. Allinstalled piping is seismically analyzed. The temporary boiler has been specified to provide the appropriate steam parameters including steam quality to perform the testing. This modification does not involve an unreviewed safety question, nor does it require a change to JAF Technical Specifications.
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ATTACHMENT I Annual Summary of Channes. Tests and Experiments for 1997/1998 JAF SE 98-016, REV. 0: DE-ENERGlZATION OF BREAKER 8 OF REACTOR Bull. DING DISTRIBUTION PANEL 71 ACA3 MODIFICATION: N/A The existing cabling for non-safety related components 02-2FIS-17A, 23A, 21 A, 26A has been identified as being susceptible to a single failure that may potentially affect other QA ,
Category I safety-related cable, specifically for the Automatic Depessurization System 1 (ADS). A postulated small break LOCA and resultant short circuit due to a failure in one of these non-safety related components, with the random single failure of the circuit breaker, could overheat the non-safety related cable and possibly cause a fire in the cable tray.
(Reference DER-98-00909).
Since the station is currently susceptible to this potential scenario, the four flow indication {
switches have been de-energized until a modification can be installed. The purpose of this evaluation is to show that temporarily opening breaker 8 of 71 ACA3 does not present the station with an unreviewed safety question.
The flow switches connected to breaker 8 of 71 ACA3 are indicating switches used to determine whether or not adequate cooling flow is provided to the RWR pump motor coolers, the RWR pump seal coolers, and whether or not abnormal RWR pump sealleakage exists. The switches have no power or control functions. Temporary de-energization of these switches does not adversely affect the operation of any equipment important to safety, Furthermore, sufficient redundant indications exist such that 02 2FIS-17A,23A, 21 A, and 26A can be de-energized without compromising the safe operation of the plant.
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This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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p ATTACHMENT I Annual Summary of Channes. Tests. and Experiments for 1997/1931 l
JAF-SE 98-017, REV. 0: SECONDARY PROTECTION FOR 71 ACA3 CIRCUlT NO. 8 l
MODIFICATION: M1 98-082 The purpose of this modification is to provide secondary protection for 71 ACA3 circuit # 8 i
with two safety-related EQ fuses. The scope of the modification is the installation of a new junction box containing two safety-related EO fuses /holdars mounted below 71 ACA3 in the Reactor Building Elevation 272'. A conduit nipple will be installed between the new junction j box and 71 ACA3 with #12AWG SIS wires. The existing cable 1RWRARC060 will be -
disconnected from circuit #8 breaker and spliced to one of the SIS wires and connect to fuse block.
The accident analysis for ADS functionality requires that the non-safety related load be j presumed to fail as a consequence of accident. The analysis requires that the ADS system i remain functionalin the presence of an additional single random failure in this case, the breaker fails to trip and ultimately the cable could damage due to fault current and may affect the ADS cables. The proposed activity will address this concern by installing two EQ fuses to provide secondary protection.
The purpose of this modification is to minimize any affect on ADS cables due to a single random failure coincident with a postulated non-safety related load failure as a consequence of the accident. The proposed modification will also address the protective device failure due to harsh environment as a consequence of an accident. The cable fed j from 71 ACA3, circuit # 8 is not required for safe shutdown. The proposed modification will '
provide backup protection to the cable 1RWRARC060 which runs in a common raceway with the ADS cables.
The proposed modification does not involve changing any set point and/or replacing any safety-related equipment. The 71 ACA3, circuit # 8 associated with this modification is not described in the Technical specification.
l Therefore, the proposed modification does not involve an unreviewed safety question, as defined in 10 CFR 50.59, based on the above analysis. The proposed activity does not degrade the Security Plan, Quality Assurance Program or the Fire Protection Program.
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ATTACHMENT I Annual Summary of Channes. Tests, and Experiments for 1997/1998 JAF-SE 98-018, REV. 0: TORUS DESLUDGE CONNECTION MODIFICATION: M1-98-081 This Nuclear Safety Evaluation documents that there is no unreviewed safety question associated with this modification. This Minor Modification M198-081 installs a hose ;
connection in the reactor building floor drain sump pump discharge line 20-1-3"-WL-151- '
3A. This hose connection will be used during the torus desludge operation to pump torus sludge to the floor drain collector tank. This modification will enhance the personnel safety and provide a cost effective way to transport the torus sludge. Torus desludge connection isolation valve 20RDW-2048 will remain closed and will be open only to pump the torus sludge. A 3 inch camlock fitting with cap will be installed downstream of the l valve 20RDW-2048. This modification does not degrade the design bases of the Liquid Radwaste System in any way that is stated in the FSAR section 11.2. This modification does not result in a change to the Technical Specifications nor does it negatively impact in any way other environmentally qualified system or structures.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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ATTACHMENT I l Annual Summary of Chanaes, Tests, and Experiments for 1997/1998 l
1 JAF-SE 98-019, REV. 0: IRM/SRM DETECTOR REPLACEMENT i 1
MODIFICATION: M198-044 i This evaluation reviews the impact of incorporating an improved design IRM/SRM detector, in conjunction with M1-98-044, IRM/SRM Detector Replacement, the improved design IRM and SRM detector assemblies have been determined to be acceptable without negatively
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impacting the design or licensing basis. The improved design utilizes the identical detector I (sensor) and connector and an improved low-loss transmission cable. The current quartz fiber-insulation of the original design transmission cable was found to be susceptible to a l low insulation resistance condition resulting in periodic signal fluctuations. The detector I cabling is being replaced with one utilizing a different insulation material and solid inner sheath versus woven materials. The new cabling has slightly improved transmission capabilities. The improved design detectors do not negatively impact the design function or method of performing the design function, and therefore this change does not represent an unreviewed safety question.
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Annual Summary of Channes. Tests, and Experiments for 1997/1998 i
JAF SE-98-021, REV. 0:
00NDENSATE STORAGE TANK (CST) THERMOSIPHON HEAT EXCHANGER PERMANENT ISOLATION MODIFICATION: N/A i l
l This evaluation addresses the design bases of the Condensate Storage Tanks, their safety 4 significance, and the effects of the elimination of the Auxiliary Boiler as a heat source. This f
safety evaluation demonstrates that the permanent isolation of the thermosiphon heat l
exchangers has no impact on plant safety and is acceptable.
It has been determined that without the addition of heat from the Auxiliary Boiler System, the CSTs can remain above 40*F for a minimum of 1.7 days under the most extreme conditions. For normal ambient winter conditions, described in the FSAR, the minimum {
l length of time for the CSTs to fall to 40 F is 6.1 days. This length of time is considered to I be sufficient for operator action to establish a recirculation path via the main condenser, i radwaste purification system, or the spent fuel pool to provide heat input, if necessary. The addition of a relatively small amount of heat from one of these sources (17 gpm @ 65 *F) will prevent the CSTs from falling below 40aF.
l The removal of the Auxiliary Boiler System as a heat source to the CSTs does not affect the safety of the plant nor are the CSTs required to mitigate the consequences of any plant accident evaluated in the FSAR or Safety Evaluation Report. Since the proposed activity neither increases the probability or consequence of an accident, an unreviewed safety question does not exist.
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ATTACHMENTI Annual Summary of Chanaes. Tests, and Experiments for 1997/1998 JAF-SE 98-022, REV.1: ORGANIZATIONAL CHANGES MODIFICATION: N/A l The organizational changes proposed by this safety evaluation involve the creation and deletion of management positions, management title changes, and the reassignment of position responsibilities and reporting relationships. These changes include:
the creation of the positions of Executive Vice President for Project Operations (EVP-PO) and Vice President - Special Activities (VP SA);
e the elimination of the positions of Senior Vice President - Power Generation (SVP-PG) and Vice President - Nuclear Operations; e a change to the title of Vice President - Engineering and Project Control (VP-E&PC) to Vice President -Engineering (VP-E);
e a change in the reporting relationship for the Vice President - Appraisal and compliance Services; e a change to the title of General Counsel - Law to Executive Vice President, 1 Secretary and General Counsel - Law; and
- a change to the title of Senior Vice President - Business Services to Executive Vice President - Business Services.
Revision 1 to this safety evaluation includes the following change:
a change in the reporting relationship for the Director - Security.
These changes reassign responsibilities and revise management titles but do not eliminate any functional requirements. The proposed changes are administrative in nature and do not involve plant equipment or operating conditions. They will not reduce the effectiveness of the management of activities or of the oversight of plant operations. Therefore, these changes do not involve an unreviewed safety question.
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ATTACHMENT I Annual Summary of Channes. Tests, and Experiments for 1997/1998 JAF-SE-98-024, REV.1: DRAINING OF THE FUEL POOL COOLING SYSTEM MODIFICATION: N/A The purpose of this safety evaluation is to determine whether any unreviewed safety question exists for the performance of Temporary Operating Procedure TOP-282, Rev. O,
" Draining of the Fuel Pool Cooling System *".
TOP-282 can essentially be broken down into five major steps. After ensuring that the Decay Heat Removal (DHR) system is in operation, the first step is to drain the Fuel Pool Cooling and Cleanup (FPCCU) system so that maintenance may be performed on the system. The second step is to lower the level in the spent fuel pool so that water level changes from DHR operation will not result in water flowing into the fuel pool cooling system while maintenance is being performed. The third step is to maintain the spent fuel pool water level while maintenance is being performed. The fourth step is to fill the spent fuel pool back up to its normal level, and the fifth step is to fill and restore to operation the FPCCU system.
There are no unreviewed safety questions with this evolution mainly because the Fuel Pool Cooling System and the Decay Heat Removal System perform no accident mitigation functions. The pool Tech Spec on level will be maintained and monitoring pool level will also be maintained constantly.
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ATTACHMENT I Annual Summary of Channes, Tests, and Experiments for 1997/1998
, JAF-SE-98-025, REV.1: HIGH PRESSURE COOLANT INJECTION AND REACTOR CORE ISOLATION COOLING SUPPRESSION POOL SUCTION STRAINER REPLACEMENT MODIFICATION: F1-98-100 Modification F198-100, Rev. O, "High Pressure Core injection and Reactor Core Isolation Cooling Suppression Pool Suction Strainer Replacement", installs new suppression pool ;
suction strainers for HPCI and RCIC system pumps. The replacement strainers are larger than the current strainers and provide additional margin for debris loading following a small or intermediate break loss-of-coolant-accident. Consistent with the original licensing basis for FitzPatrick, the replacement strainers have been evaluated to ensure adequate NPSH l would be available to the pumps with the suction strainers 50% blocked. The strainer assemblies have been analyzed and qualified for seismic md Mark I hydrodynamic loads i specified in the FSAR and Plant Unique Analysis Report. l l
A review of the modification in accordance with 10CFR50.59 concluded that the modification does not increase the probability or consequences of an accident or of a malfunction of equipment important to safety previously evaluated in the safety analysis report. Further, the possibility of an accident or malfunction of a different type than any !
evaluated previously la the safety analysis report will not be created. The margin of safety l l' as defined in the Technical Specification Bases is not reduced and no Technical l l
Specification change is required. This change, therefore, does not involve an unreviewed safety question.
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ATTACHMENT I Annual Summary of Chances. Tests. and Experiments for 1997/1998 JAF-SE-98-026, REV.1: ELECTROCHEMICAL POTENTIAL (ECP) MONITORING / RETIRE CRACK ARREST VERIFICATION (CAV) SYSTEM / RELOCATE DISSOLVED GAS CALIBRATORS MODIFICATION: M197-132 i
Modification M1-97-132: (1) installs an Electrochemical Potential (ECP) flange assembly at f the existing decontamination flange connection located between Reactor Water I Recirculation (RWR) system pump 02-2P-1B and recirculation loop B pump suction isolation valve 02-2MOV-43B and upgrades the existing incore Stress Corrosion Monitor Data Acquisition System with a new computer and software. The ECP flange assembly is being installed to provide the capability to collect data to evaluate the extent to which primary system chemistry conditions will resist or mitigate intergranular stress corrosion cracking (IGSCC); (2) Retire the Crack Arrest Verification (CAV) system. Operating experience has i shown that CAV data is not representative of reactor recirculation coolant conditions with respect to electrochemical potential; and (3) Relocate Dissolved Gas Calibrators 7A and 78, from inside panel 95SP-7 to the outside wall of the panel to reduce occupational radiation exposure.
This modification does not involve an unreviewed safety question nor does it require a change to JAF Technical Specifications.
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ATTACHMENT I Annual Summary of Channes, Tests, and Experiments for 1997/1998 JAF SE-98-027, REV. 0: COLLECTION OF HPCI TORUS SUCTION ISOLATION VALVE l LEAK DATA MODIFICATION: N/A The purpose of this safety evaluation is to determine whether any unreviewed safety question exists for the performance of Temporary Operating Procedure TOP-281, Rev. O,
" Collecting HPCI Valve Leakage Data *". TOP-281 can essentially be broken down into three major steps.
- Measurement of static pressure between 23MOV-57/58 Collection of baseline leakage from between 23MOV-57/58 - prior to cycling valves e
Cycling of valves individually and then collecting leakage to obtain additional data Collecting leakage will require that the pressure boundary between 23MOV-57 and 23MOV-58 is broken and temporary equipment installed to collect the leakage. There are no unreviewed safety questions with this evolution because the HPCI and Primary Containment isolation systems will remain capable of performing their required functions throughout the duration of the activity.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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ATTACHMENT I Annual Summary of Chanoes. Tests, and Experiments for 1997/1998 i JAF-SE 98-030, REV. 0: INSTALLATION OF TEMPORARY MANUAL VALVE ON TIP i
1 PENETRATION MODIFICATION: TEMPORARY MODIFICATION 98-052 The following actions are proposed in the event that a TIP ball valve cannot be locally l closed manually:
Disconnect the tubing from the shear valve to the chamber shield at the shear valve coupling and connect a capped manual valve that has been leak-rate tested to the shear valve coupling.
l Leak rate test the connections to the manual valve via the cap.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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ATTACHMENT I l Annual Summary of Chances. Tests. and Experiments for 1997/1998 JAF-SE 98-032, REV. 0: RPV SHELL WELD EXAMINATION WITH GERIS 2000 ID EQUIPMENT MODIFICATION: N/A The present FitzPatrick ISI 10-year interval requirement, per the committed ASME XI 1989 Edition, will be carried out according to the FitzPatrick Plant ISI Program [[::JAF-ISI-0002|JAF-ISI-0002]],
" Third inservice inspection Interval Inspection Program". The GE Remote Ultrasonic Manipulator System [GERIS 2000] will be utilized for the required examinations. The mechanical computer controlled manipulator system, coupled with a GENE designed Ultrasonic Test (UT) system has the capability to conduct UT inspection of the inner surface of BWR reactor pressure vessels above the shroud support plate.
The lifting equipment associated with the GERIS 2000 is designed to meet the requirements of NUREG 0612, such that a mechanical failure, that would allow the j equipment to be dropped during installation or removal, is not considered credible. The movement of the GERIS 2000 equipment over the refueling floor and the RPV will follow l approved pathways. The GERIS 2000 will have no effect on normal refueling activities.
1 in evaluating the operation of the GERIS 2000 at the FitzPatrick Plant, the following considerations were made within the context of 10CFR50.59: 1) heavy loads analysis 2) seismic analysis; 3) lost parts assessment: 4) radiological review; 5) fire protection; and i
- 6) material compatibility. Fuel movement is prohibited during GERIS 2000 installation and removal. Based on the safety evaluation performed, the GERIS 2000 equipment and plant procedures, it is concluded that the ISI can be safely conducted end will not result in an unreviewed safety question.
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ATTACHMENT l Annual Summarv of Cham , Tests, and F=== iments for 1997/1998 1
JAF SE-98-033, REV.1: REPLACEMENT OF 13MOV-16 L MODIFICATION: M1-97-033 The existing Reactor Core Isolation Cooling (RCIC) System Turbine Steam Supply Primary Containment Outboard Isolation Valve (13MOV-16) has been susceptible to failure of 10CFR50, Appendix J, Type C local leak rate tests performed in accordance with the
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James A. Fitzpatrick Primary Containment Leakage Ratt. Testing Program. The repeat
. failures have been attributed to corrosion scale build-up on the valve seating surfaces and to seat scoring -
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i The existing gate valve is being replaced with a gate valve of a different style and supplied by a different valve manufacturer. In addition, an auxiliary, small piping arrangement will be attached to the valve which will include manual valves. This will be configured to provide a bonnet overpressure bypass during normal plant operation and two sets of double isolation valves for localized test and drain connections to support an alternate means of local leak rate testing. The existing motor operator will be reinstalled on the replacement isolation valve. The purpose of this evaluation is to show that the replacements of valve 13MOV-16 and the auxiliary piping arrangement do not present an unreviewed safety question.
This valve application is within the scope of valves evaluated in response to NRC Generic i Letter 89-10 and Generic Letter 95-07. Closing thrust requirements for the replacement valve were determined using the EPRI MOV Performance Prediction Methodology and the i established requirements of the JAF MOV Program. Actuator output capability was d determined to satisfy the operational requirements under design basis conditions and satisfy the recommendations of GL 89-10. With regard to GL 95 07 concerning gate valve pressure locking and thermal binding, the bonnet overpressure bypass capability is retained by the new installation. This assures that a pressure locking condition cannot occur.
Thermal binding is not a concern for this valve since it is normally open during plant power -
operation. It is periodically closed to perform a stroke time test, but the duration of it's '
l . being in a closed state is sufficiently brief so as not to cause thermal binding.
l The new 13MOV-16 valve is being supplied in accordance with the ASME B&PV Code, Section lli design and construction criteria for Class 1 valves. The activities to replace this valve will occur during a plant refueling outage. The proposed modification review process has performed the requisite analyses for component and piping stress and seismic loading and for motor operator sizing. Adequate margin of safety has been determined to exist for the plant normal and potential accident operational requirements ;
or the affected components. Therefore, the replacement of the 13MOV-16 valve does
!- not adversely affect the operational capability of the valve, compromise the safe '
l operation of the plant or introduce an unreviewed safety question.
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ATTACHMENT I Annual Summary of Chanaes. Tests, and Experinmnts for 1997/1998 JAF-SE-98-034, REV. 0:
EMERGENCY SERVICE WATER (ESW) SYSTEM CHECK VALVE REPLACEMENT MODIFICATION: M1-97-026 Minor Modification M1-97-026, " Emergency Service Water (ESW) System Check Valve Replacement" will add stainless steel globe stop check valves 46ESW-40A (ESW Keep Full Loop A Check Valve) and 46ESW-40B (ESW Keep Full Loop B Check Valve) as replacements for stainless steel piston lift check valves 46ESW-40A and 46ESW-408 in the Emergency Serrice Water (ESW) System. The Keep Full system, including these check valves, enhances toe operability of the ESW system by ensuring that ESW piping is full of water and pressurized with Service Water. This will preclude water hammer events in this piping. The proposed modification will provide the capability to manually seat 46ESW-40A and 46ESW-408 periodically to clear any residue / debris that may have collected in the valves. This action ensures maximum potential to perform their design function as validated via Surveillance Test Procedure ST-8R, " Emergency Service Water Check Valve and Strainer Test". The ESW is classified as a nuclear safety-related system. The safety objective of the Emergency Service Water System is to provide cooling to Emergency Core Cooling System components and other equipment essential to safe reactor shutdown following a design basis LOCA. Therefore, check valves 46ESW-40A and 46ESW-40B are classified as Quality Assurance (QA) Category i since the valves constitute the boundary between the Service Water (SWS) and the ESW systems. The changes proposed by this modification do not degrade the Security Plan, Quality Assurance Program, the Fire Protection Program, the Environmental Report or the Emergency Plan. This change does not constitute an Unreviewed Safety Question per 10CFR50.59.
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ATTACHMENT I Annual Summary of Chanoes. Tests, and Experiments for 1997/1998 JAF-SE-98-035, REV.1: TORUS DEWATERING MODIFICATION: TEMPORARY MODIFICATION 98-059A De-watering, filtration / purification, water storage and reflooding of the James A. FitzPatrick Nuclear Plant torus during RFO-13 is necessary to provide torus access and a radiologically safe environment to support the installation of replacement Emergency Core Cooling (ECCS) and RCIC Strainers (modifications F1-97-031 and F1-98-100). This proposed activity can be accomplished without impairing primary or secondary containment control per the plant Technical Specifications and is consistent with the various design documents referenced in section 3.0 above.
The radiological consequences of postulated liquid leakage from the torus water storage tank is bounded by the existing UFSAR analysis. Nevertheless, this tank is equipped with a secondary containment, floating cover, leak detection and sample points for radiological activity testing. In addition, transfer hose, pumps and filtration / purification apparatus used during processing operations will have secondary containment provisions or catch basins to reduce the potential for release of torus effluent.
Following re-fill of the torus, the ECC systems will be administratively controlled, as necessary, to prevent injection of cold water (< 68 F) to the reactor vessel, This will ensure that the water has reached sufficient temperature to prevent any reactor vessel nil -
ductility, positive reactivity or shutdown margin concerns.
The temporary system interfaces with permanent NSR systems only and has no adverse impact on any safety-related system.
Based on the above, torus de watering, filtration / purification, water storage and reflooding of the James A. FitzPatrick Nuclear Plant torus does not represent an unreviewed safety question and is consistent with the plant design. ,
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ATTACHMENT I Annual Summary of Chanaes. Tests, and Experiments for 1997/1998 JAF-SE-98-036, REV. 0:
LIFTING POWER LEADS FOR AIRLOCK POSITION INDICATION SWITCHES MODIFICATION: TEMPORARY MODIFICATION 98-062 i The existing circuit for non-safety related components 16PNS-1 A,18,1C,1D, 2A &
2C has been identified as being susceptible to a single failure that may potentially affect electrical penetrations 16X-100F and 16X-1018. In the event of a LOCA, two (2) of these non-safety related loads (1 A and 2A only) must be presumed to fail and to piace faults on their circuits, if, as is required by the single failure criterion, their one level of protection is also postulated to fail, then there is no overload protection remaining on the circuits and the faults will not be cleared. Therefore, the penetration through which these circuits pass could be damaged and containment integrity could be violated. l I
Sinca the station is currently susceptible to this potential scenario, the position indicating switches associated with non-safety related leads will be lifted during power operations until a modification is installed.
The position indication switches provide the status of the airlock Access and Escape Hatch door position remotely. The temporary de-energizing of the components does not adversely affect the operation of any equipment important to safety. Furthermore, sufficient administrative controls will exist such that the limit switches and their indicating lights can be de-energized without compromising the safe operation of the plant.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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ATTACHMENT I e_nfael Summary of Channes, Tests, and Experiments for 1997/1998 JAF-SE-98-037, REV. 0: DE-ENERGlZING BREAKER 3 OF REACTOR BUILDING DISTRIBUTION PANEL 71 ACB3 MODIFICATION: TEMPORARY MODIFICATION 98-061 The existing cabling for non-safety related components 02-2FIS-17B 23B, 218,26B has been identified as being susceptible to a single failure that may potentially effect other QA Category I equipment (electrical containment penetration). With a postulated LOCA and resultant short circuit due to a failure in one of these non-safety related components, with the random single failure of the circuit breaker, could overheat the electrical containment penetration. (Re: DER-98-01910).
Since the station is currently susceptible to this potential scenario, the four flow indication switches will be de-energized, until a fuse modification can be installed. The purpose of this evaluation is to show that temporarily opening breaker 3 of 71 ACB3 does not result in an unreviewed safety question.
The flow switcher connected to breaker 3 of 71 ACB3 are indicating switches used to determine whether or not adequate cooling flow is provided to the RWR pump motor coolers, the RWR pump seal coolers, and whether or not abnormal RWR pump seal leakage exists. The switches have no power or control functions. Temporary de-energization of these switches does not adversely affect the operation of any equipment important to safety. Furthermore, sufficient redundant indications exist such that 02-2FIS-17B,23B, 21B, and 268 can be de-energized without compromising the safe operation of the plant.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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Annual Summary of Chanaes Tests, and Experiments for 1997/1998 JAF-SE 98-038, REV. 0:
MSR DRAIN TANK TRIP LOGIC IMPROVEMENT i
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MODIFICATION: M1-96-095 I
JAF Minor Modification No. Mi-96-095, MSR Drain Tank Trip Logic Improvement, changes i
I the moisture separator high level turbine trip logic to reduce the probability of turbine trips and consequent reactor scrams caused by single component failures. This modification protects against single component failure and does not prevent the necessary automatic actions or operator actions in the event of a moisture separator high level trip. This modification does not affect any safety-related components, systems or interfaces, and the reduced probability of spurious trips precludes unnecessary challenges to reactor protection systems and plant operators and equipment.
No changes to the FSAR or Technical Specifications result from this modification. This modification does not affect the operation, function or method of performing the function of any safety or important to safety equipment described in the SAR. This modification is installed and tested in accordance with existing and approved plant procedures and does not constitute a test or experiment not described in the SAR.
This modification does not present an unreviewed safety question pursuant to 10CFR50.59, since this modification: (i) does not affect the function, operation, design requirements, test requirements, or qualifications of safety-related equ!pment or equipment important to safety as described in the SAR, (ii) does not create any new failures and is within the existing design bases as described in the SAR, and (iii) does not affect design or operating parameters, test or surveillance requirements, safety limits, limiting safety system settings, or limiting conditions of operation ss defined in the basis of any technical specification, i
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ATTACHMENT l !
Annual Summary of Channes, Tests, and Fxperiments for 1997/1998 l l t
! JAF-SE-98-040, REV. 0:
TEMPORARY DE-ENERGlZATION OF DRYWELL COOLING l MOTOR OPERATED VALVE (MOV) POWER SUPPLIES !
l MODIFICATION: TEMPORARY MODIFICATION 98-063 l I The existing circuits for non-safety related components 15MOV-100A1/A2/A3/A4/B1/B2/ I B3/84 have been identified as being susceptible to a single failure that may potentially affect QA Category I safety-related containment penetrations. A postulated short circuit due to a failure of one of these non-safety related valves, with the single failure of the circuit breaker in the same circuit, could overheat the cable and possibly cause damage to the electrical penetration. (Reference DER-98-01910).
j Since the station is currently susceptible to this potential scenario, the eight MOVs will be de-energized until a modification can be installed or the configuration has been determined to be acceptable. The purpose of this evaluation is to show that temporarily opening breakers that supply power to these valves does not present the station with an j unreviewed safety question.
The MOVs and their associated indicating switches are used to provide cooling water flow and flow isolation to the individual cooling coils on the non-safety-related drywell coolers.
Temporary de-energization of these components does not adversely affect the operation of any equipment important to safety. Furthermore, sufficient redundant indications exist such that the MOVs and their indicating lights can be de-energized without compromising the I' safe operation of the plant.
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ATTACHMENT I Annual Summary of Channes, Tests, and Experiments for 1997/1998 JAF SE-98-042, REV. 0:
HIGH PRESSURE COOLANT INJECTION (HPCI) AND REACTOR CORE ISOLATION COOLING (RCIC) POST MODIFICATION TESTING MODIFICATION: N/A Preoperational tests have been developed to verify design flow capability for the HPCI and RCIC pumps using new suppression pool suction strainers installed during Refueling Outage
- 13. To provide a flow path capable of handling full pump flow, a valve alignment is established to allow water from the suppression pool to be pumped to the Condensate Storage Tanks. The test procedures provide details for temporary changes to control circuits for motor operated valves in the system flow paths and limitations to preclude operating outside Technical Specification limits for suppression pool level.
A review of the tests in accordance with 10CFR50.59 concluded that the prescribed testing does not increase the probability or consequences of an accident or of a malfunction of equipment important to safety previously evaluated in the safety analysis report. Further, the possibility of an accident or malfunction of a different type than any evaluated previously in the safety analysis report will not be created. The margin of safety as defined in the Technical Specification Bases is not reduced and no Technical Specification change is required. Performance of these tests, therefore, does not involve an unreviewed safety question.
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ATTACHMENT I Annual Summary of Chanaes, Tests, and Experiments for 1997/1998 [[::JAF-SE-98|JAF-SE-98]] 044, REV. 0:
INSTALLATION OF A VALVE IN 33AOV-D-2 (RETIRED)
POSITION FOR OUTAGE WATER MOVES MODIFICATION: TEMPORARY MODIFICATION 98-072 A manual valve will replace a spectacle flange in a location that had a manually operated AOV in the original plant design. This accessible valve will be used to support outage water moves.
The proposed activity more closely represents the original design condition and will not !
increase probability or consequences of adverse situations. I The proposed activity has no effect on the various other programs currently being !
evaluated under this process. None of the above actions results in an unreviewed safety question.
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ATTACHMENTI Annual Summarv of Chanaes. Tests, and Experiments for 1997/199,8 l
JAF-SE-98-045, REV. 0:
DISASSEMBLY OF THE REACTOR PRESSURE VESSEL (RPV)
HEAD STRONGBACK CAROUSEL MODIFICATION: N/A l The purpose of this Nuclear Safety Evaluation is to demonstrate the acceptability of lifting ,
the new RPV head strongback carousel assembly by the Reactor Building crane. This activity is necessary to facilitate the disassembly of the component to allow its removal from the Refuel Floor prior to RFO13.
1 The movement of the RPV Head Strongback Carousel on the Refuel Floor, using the l Reactor Building Crane Main Hook, and the subsequent disassembly of the unit as i evaluated in this NSE meets all the applicable requirements of NUREG 0612 for heavy loads. The evaluation demonstrates that the activity does not constitute any unreviewed safety question as defined by 10CFR50.59 nor does it involve significant hazards as defined in 10 CFR 50.92. Therefore, the activity can be performed as proposed.
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fe ATTACHMENT I Annual Summary of Chanm, Tests, and F=naciments for 1997/1998 i.
JAF-SE-98-046, REV.1: JAF ORGANIZATION CHANGE l
l MODIFICATION: N/A i
l The organizationel change proposed by this NSE involves the reassignment of reporting responsibilities for the position of Training Manager from the General Manager Support Services directly to the Site Executive Office, thus changing from five to six the number of managers and senior managers reporting directly to the SEO.
This reassignment of responsibility does not eliminate any function'or Training Program requirement. The proposed change is administrative in nature and does not involve plant i
equipment or operating condition. The change does not reduce the effectiveness of the management of activities or oversight of plant operations. Therefore, the change does not involve an unreviewed safety question.
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f ATTACHMENT I An iual Summary of Channes, Tests, and Experiments for 1997/1998 JAF SE-98-047, REV. 0: MODIFY 23MOV-14 TO ELIMINATE LEAKAGE MODIFICATION: M1-98-146 The purpose of this proposed activity is to modify the internals of the HPCI steam admission valve,23MOV-14, to prevent leakage from the valve. Also, the closure logic is modified to close the valve on position rather than torque. This modification has no adverse impact on the safety-related function of the valve, which is to open. Also, the torque switch is left active and in the logic to preclude valve damage should the position switch fail. A body drain is added which will drain the valve body to an existing condensate pot and drain system. The implementation of this activity will not cause any accidents or malfunction of equipment important to safety as described in the JAF FSAR. The addition of a few additional feet of small bore piping to the plant will not significantly increase the possibility of a line break accident at JAF. In addition, the modifications to 23MOV-14 does not increase the radiological consequences of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR nor can it cause any new consequences of an accident or malfunction of equipment important to safety. The proposed activities associated with the 23MOV-14 is performed to preclude valve leakage and does not adversely affect the HPCI System and therefore the bases for the Technical Specifications because the new valve components will be upgraded to a more reliable and leaktight design. Also, the forces required to open and close the valve will be reduced. In conclusion, the modification of the 23MOV-14 and the addition of the body drain and the update of the FSAR to reflect the addition of the body drain is acceptable since this activity does not constitute an unreviewed safety question pursuant to 10CFR50.59.
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ATTACHMENT I Annual Summary of Chanaes. Tests and Experiments for 1997/1998 l
l l JAF-SE-98-048, REV. 0: REPLACEMENT OF 39FCV-110 - JUMPER TO MAINTAIN INSTRUMENT AIR q
i MODIFICATION: TEMPORARY MODIFICATION 98 075 i Temp Mod 98-075 will install a temporary piping to maintain instrument air and breathing air while 39FCV-110 is being replaced and 39SAS-2D is being repaired installation of the temporary piping and isolation of the piping around 39FCV-110 will remove a small portion of the compressed air piping and the entire Service Air System from service. Instrument air >
l and breathing air supply will not be interrupted. The Service Air System will be repressurized while the work is being performed which is estimated to take two shifts but I may take up to two days.
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Service air is not involved in any accident initiation scenario and would not increase the I consequences of an accident. The Service Air System does not perform any safety-related functions and can not cause a malfunction of any safety-related equipment. The Service Air System is not listed in the Technical Specifications, therefore, it can not reduce the margin of safety. None of the above actions results in an unreviewed safety question.
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7 ATTACHMENT I Annual Summary of Chane=s Tests, and Experiments for 1997/1998 JAF-SE 98-050, REV. 0: USE OF RCIC MANUAL INITIATION PUSHBUTTON AS ALLOWED BY OPERATING PROCEDURES (ops), EMERGENCY OPERATING PROCEDURES (EOPs) AND EOP SUPPORT PROCEDURES MODIFICATION: N/A '
This s3fety evaluation determined no unreviewed safety question exists for using the RCIC meaual initiation pushbutton installed per modification F1-92-173. The pushbutton may be used to either facilitate actions required by the EOPs or to allow proactive use of RCIC prior to the RPV level initiation signal being received. The pushbutton switch was originally installed to ensure that a fire would not preclude RCIC remote start capability. The modification safety evaluation did not address the use of this switch for occasions other than during plant fires and thus site procedures governing these other situations were not revised..The reason for this safety evaluation is to clarify what the requirements are for use of this pushbutton for events other than plant fires.
The switch will be used to carry out desired actions to mitigate the consequences of the
. situation at hand, based on plant conditions. The ability to manually initiate RCIC, provided initiation or shutdown signals do not exist is an original design feature of the RCIC system.
The use of a single manual pushbutton switch is functionally the same as using multiple operator actions per existing procedures. This improves the ability of plant operators to respond to conditions which require the use of the RCIC system in a timely manner.
This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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i ATTACHMENT I Annual Summary of Chanaes, Tests, and Experiments for 1997/1998
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JAF-SE 98-051, REV. 0:
INBOARD MAIN STEAM ISOLATION VALVES (MSIVs) NAMCO CONNECTORS ]
MODIFICATQON: N/A
)
The proposed activity being evaluated consists of accepting the installation as is for the plug-in cable supplied with Namco EC210 Shielded Cable Seal / Connector Assemblies for the MSIV SVCA, located inside the primary containment. Approximately less than 3 feet of the plug-in cables are not in conduit. UFSAR Section 7.1.9.a states that all cables inside containment are installed in conduit.
An unreviewed safety question is not involved because the plug-in cables installed in free air are environmentally qualified and the use of conduit for the subject installation is not required to assure the functionality of the MSIVs. Additionally, no requirements exist in the JAF Electrical Separation Criteria,10CFR50 Appendix R or in the Accident Analysis that requires the use of metallic conduit for the subject installation.
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ATTACHMENT I Annual Summary of Chances Tests, and Experiments for 1997/1998 I
JAF-SE-98-052, REV. 0: EVALUATION OF DISABLING 10MOV-21 A AND B IN THE l CLOSED POSITION MODIFICATION: TEMPORARY MODIFICATION 98-120 RHR Heat Exchanger Discharge to Torus isolation Valves 10MOV-21 A and 10MOV-21B will be disabled in the closed position by removing the overload heaters from their respective controllers to provide isolation of their respective lines from the Torus when Primary Containment Integrity is tr;c;uired. Following failure of these valves to meet differential pressure operability requirements they were descoped from the JAF Generic Letter 8910 program on the basis that they would be isolated in the closed position.
The safety function of these normally closed valves is to close and remain closed when required for containment isolation. Disabling them in the closed position satisfies the requirements for Primary Containment integrity as set forth in the UFSAR and Technical j Specifications. The non-safety related functions of these valves are to open and close as j required to support operation of the RHR system in the Steam Condensing Mode. The Steam Condensing Mode is not safety-related, and is not credited in the accident analysis or in the Appendix R safe shutdown analysis. Therefore disabling these valves in the closed i
' position does not constitute an unreviewed safety question.
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ATTACHMENT I Annual Summarv of Chama. Tests, and F=Mments for 1997/1000 i
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' JAF-SE-98-053, REV. 0:
EMERGENCY OPERATING PROCEDURE (EOP) REVISION TO IMPLEMENT BWROG EMERGENCY PROCEDURE AND SEVERE ACCIDENT GUIDELINES, REVISION 1 MODIFICATION: N/A l
l This revision of the JAF Emergency Operating Procedures (EOPs) implements the Boiling !
Water Reactor Owner's Group Emergency Procedure and Severe Accident Guidelines, j Revision 1. The EOPs were revised to incorporate industry developed strategies for ATWS/ stability and severe accident management. The EOP steps revised for ATWS/ stability were in accordance with the guidance approved by the NRC SER on l ]'
Modifications To The BWR Emergency Procedure Guidelines To Address Reactor Core Instability. Incorporating the severe accident management guidance required removing strategies from the EOPs which were applicable only to severe accident conditions and placing them in Severe Accident Operating Guidelines (SAOGs). Transition steps directing the operating crew to exit the all EOPs and enter the SAOGs were included in the EOP revision. All of the EOP steps to transition to the SAOGs occur when conditions have l degraded beyond the plant design and licensing bases.
. A review of the EOP changes in accordance with 10CFR50.59 concluded that they do not increase the probability or consequences of an accident or malfunction of equipment
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j important to safety previously evaluated in the safety analysis report. The EOP changes do not create the possibility of an accident or equipment malfunction of different than any previously evaluated in the safety analysis report. The margin of safety as defined in the Technical Specifications is not reduced and no Technical Specification change is required. ;
j Tlierefore, this revision of the JAF EOPs does not involve an unreviewed safety question. !
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ATTACHMENT I Annual Summary of Chanaes, Tests, and Experiments for 1997/1998 JAF SE-98-054, REV. 0: TRAVERSING IN CORE PROBE (TIP) SYSTEM Na PURGE MODIFICATION: N/A The proposed activity is to connect a N2source and check valve to the TIP System downstrearn (away from containment) from the ball and shear valves. Moving the drive wheel position encoder to a positive number opening the TIP ball valve. Nitrogen pressure will be maintained at less than 20 psig. If during the evolution a containment i isolation signal were received, the ball valve would close per the normal isclation logic.
1 This safety evaluation concluded that this activity does not constitute an unreviewed safety question pursuant to 10 CFR 50.59.
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