Text

Annual Operating Rept for Browns Ferry Nuclear Plant for 1990. W/910301 Ltr
	ML18033B648
Person / Time
Site:	Browns Ferry
Issue date:	12/31/1990
From:	Carier P; TENNESSEE VALLEY AUTHORITY
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	NUDOCS 9103060307
	Download: ML18033B648 (255)
	v • d • e

ANNUAL OPERATING REPORT BROWNS FERRY NUCLEAR PLANT TENNESSEE VALLEY AUTHORITY JANUARY-DECEMBER 1990 DOCKET NUMBERS 50-259, 50-260, AND 50-296 LICENSE NUMBERS DPR-33, DPR-52, AND DPR-68 9103060307

TABLE OF CONTENTS Page Operational Summary . ~ ~ ~ 2 Last of Acronyms ~ ~ ~ 6 10 CFR 50.59(b)(2) 'Summary of Safety Evaluations Final Safety Analysis Report Changes. . . . . . . . . . . . . 9 New Instructions. ~ ~ ~ ~ ~ 32 Procedure Revisions ~ ~ ~ ~ ~ ~ ~ ~ 40 Special Operating Conditions. ~ ~ ~, 45 Special Tests ~ ~ ~ 51 Temporary Alterations ~ ~ ~ 57 Plant Modifications . . . . . . . . . . . . . . . . . . . . . 60 Regulatory Guide 1.16, Section l.b.(4) 1990 Release Summary. . . . . . . . . . . . . . . . . . . . . 143 Technical Specification 6.9.1.2 1990 Occupational Exposure Data . 144 Challenges to or Failures of Main Steam Relief Valve s ~ 146 Technical Specification 6.9.2.1 Reactor Vessel Fatigue Usage Evaluation . 147 Technical Specification Appendix B, 3.2.2 Transmission Line Corridor Herbicide Usage. ~ ~ ~ ~ ~ 148

OPERATIONAL

SUMMARY

JANUARY-DECEMBER 1990 Unit remains on administrative hold to resolve various TVA and NRC concerns.

Unit remains on administrative hold to resolve various TVA and NRC concerns.

Modifications, operation, and maintenance 'work continues to support restart.

Outage work was performed in the following areas:

1. Environmental qualification and electrical issue modifications 2~ Appendix R modifications

3. Drywell structural steel modifications 4~ Control Room habitability modifications

5. Diesel air start system modifications

6. Fire Protection system upgrades

7. Main Steam Line Radiation upgrade

8. Radwaste System upgrades

9. MSIV control air hose upgrade

10. ADS upgrades

11. SGTS upgrades

12. SPDS installation

13. Reactor building and turbine building CAN upgrades

14. PASS modifications

15. Offgas system upgrades LlHX~

Unit, remains on administrative hold to resolve various TVA and NRC concerns.

I

RATING DATA REPORT Ki2RKZIE PER T

!Notes

l. Unit Name:

2. Reporting Period: D b

3. Licensed Thermal Power (Wt):

4. Nameplate Rating (Gross We):

5. Design Electrical Rating (Net We):

6. Maximum Dependable Capacity (Gross We):

7. 'aximum Dependable Capacity (Net We):

8. If Changes Occur in Capacity Ratings (Items Number 3 Through 7) Since Last Report, Give Reasons:

9. Power Level Tb Which Restricted, If Any (Net We):

10. Reasons For Restrictions, If Any:

This Month Yr-to-Date Cumul ati ve 12 'umber of Hours Reactor Has Critical

13. Reactor Reserve Shutdown Hours

14. Hours Generator On-Line

15. Unit Reserve Shutdown Hours

16. Gross Thermal Energy Generated (WH)

17. Gross Electrical Energy Generated (WH)

18. Net Electrical Energy Generated (WH)

19. Unit Service Factor 20 'nit Availability Factor

21. Unit Capacity Factor (Using HDC Net) 22, Unit Capacity Factor (Using DER Net)

23. Unit Forced Outage Rate

24. Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration of Each):

/

OPERATING DATA REPORT DOCKET NO.

PREPARED BY TELEPHONE

.!Notes

1. Unit Name: w

2. Reporting Period:

3. Licensed Thermal Power (HWt):

4. Nameplate Rating (Gross HWe):

5. Design Electrical Rating (Net HWe):

6. Haximum Dependable Capacity (Gross HMe):

7. Maximum Dependable Capacity (Net HWe):

8. If Changes Occur in Capacity Ratings (Items Number 3 Through 7) Since Last Report, Give Reasons:

9. Power Level To Which Restricted, If Any (Net We):

10. Reasons For Restrictions, If Any:

,This Honth Yr-to-Date Cumul ati ve

12. Number of Hours Reactor Was Critical

13. Reactor Reserve Shutdown Hours

14. Hours Generator On-Line

15. Unit Reserve Shutdown Hours 16.,Gross Thermal Energy Generated (HWH)

17. Gross Electrical Energy Generated (HMH)

18. Net Electrical Energy Generated (HWH)

19. Unit Service Factor

20. Unit Availability Factor

21. Unit Capacity Factor (Using HDC Net)

22. Unit Capacity Factor (Using DER Net)

23. Unit Forced Outage Rate

24. Shutdowns Scheduled Over Next 6 Honths (Type, Date, and Duration of Each):

0 0

OPERATING DATA REPORT DOCKET NO.

PREPARED BY TELEPHONE P N T

.!Notes I

1. Unit Name:

2. Reporting Period:

3. Licensed Thermal Power (HMt):

4. Nameplate Rating (Gross HWe):

5. Design Electrical Rating (Net HWe):

6. Haximum Dependable Capacity (Gross HWe):

7. Haximum Dependable Capacity (Net HWe):

8. If Changes Occur in Capacity Ratings (Items Number 3 Through 7) Since Last Report, Give Reasons:

9. Power Level To 'Hhich Restricted, If Any (Net HMe):

10. Reasons For Restrictions, If Any:

This Honth Yr-to-Date Cumulative

11. Hours in Reporting Period

12. Number of Hours Reactor Has Critical

13. Reactor Reserve Shutdown Hours

14. Hours Generator On-Line

15. Unit Reserve Shutdown Hours

16. Gross Thermal Energy Generated (HWH)

17. Gross Electrical Energy Generated (HWH)

18. Net Electrical Energy Generated (HWH)

19. Unit Service Factor .4

20. Unit Availability Factor

21. Unit Capacity Factor (Using HDC Net)

23. Unit Forced Outage Rate

24. Shutdowns Scheduled Over Next 6 Honths (Type, Date, and Duration of Each):

0449k/10

LIST OF ACRONYMS ACI American Concrete Institute ACU Air Conditioning Unit ADS Automatic Depressurization System AEC Atomic Energy Commission AHU Air Handling Unit ALARA As Low As Reasonably Achievable-AOT Anticipated Operating Transient ARI Alternate Rod Insertion ARM Area Radiation Monitor ASME American Society of Mechanical Engineers ATU - Analog Trip Unit AUO Assistant Unit Operator BFN Browns Ferry Nuclear Plant BWROG Boiling Water Reactor Owners Group CAD Containment Atmosphere Dilution Continuous Air Monitor CAS Control Air System CASA Common Accident Signal A CASB Common Accident Signal B CAQR Condition Adverse to Quality Report CB Circuit Breaker CBV Control Bay Ventilation CCM Component Cooling Water CDWS Condensate and Demineralized Water System CEB Civil Engineering Branch CLIP Core Limits Program CPT Control Power Transformer CRD Control Rod Drive CRLD Change Request Licensing Document CREV Control Room Emergency Ventilation CS Core Spray Cooling Mater DBA Design Basis Accident DBVP Design Baseline Verification Program DC Direct Current DCRM Document Control Records Management DCA Drywell Control Air DCN Design Change Notice DCR Design Change Request DD Drawing Discrepancy DG Diesel Generator DSAS Diesel Start Air System ECCS Emergency Core Cooling System ECN Engineering Change Notice ECSA Electrical Conductor Seal Assembly EECW Emergency Equipment Cooling Water EPG Emergency Procedure Guidelines ESF Engineered Safeguards Feature FHA Fuel Handling Accident FIT Flow Indicating Transmitter FLC Fuel Loading Chamber 0570S

0 LIST OF ACRONYNS FPCGCS Fuel Pool Cooling and Cleanup System Flow Recorder FSAR Final Safety Analysis Report FSV Flow Solenoid Valve FW Feedwater GE General Electric HCV Hand Control Valve HELB High Energy Line Break HPCI High Pressure Coolant Injection HPFP High Pressure Fire Protection HVAC Heating, Ventilation and Air Conditioning ILRT Integrated Leak Rate Test IRM Intermediate Range Monitor JB Junction Box LCO Limiting Condition for Operation LLRW Low Level Radwaste LOCA Loss of Coolant Accident LPCI Low Pressure Coolant Injection LPCS Low Pressure Core Spray LRM Logarithmic Radiation Monitor LTPU Long Time Pick-up MCC Motor Control Center MCFL Maximum Combined Flow Limiter MCPR Minimum Critical Power Ratio MCR Main Control Room MG Motor Generator MOV Motor Operated Valve MS Main Steam MSIV Main Steam Isolation Valve MSL Main Steam Line MSRV Main Steam Relief Valve NE Nuclear Engineering OPC Outside Primary Containment OV Over Voltage PASS Post Accident Sampling System PATH Power and Thermal Limits Program PCIS Primary Containment Isolation System PCLD Primary Coolant Leak Detection PORC Plant Operations Review Committee PRFO Pressure Regulator Failure Open PSI Pounds Per Square Inch PSIG Pounds Per Square Inch Gravity PSPDS Phase I Safety Parameter Display System PSS Public Safety Security PSTG Plant Specific Technical Guidelines PRFO Pressure Regulator Failure Open QA Quality Assurance RADCON Radiological Controls [Organization]

RBCCW Reactor Building Closed Cooling Water RCIC Reactor Core Isolation Cooling RCW Raw, Cooling Water RF Radio Frequency RFW Reactor Feedwater 0570S

LIST OF ACRONYMS RHR Residual Heat Removal RHRS Residual Heat Removal System RHRSW Residual Heat Removal Service Water RMCS Rod Manual Control System RMOV Reactor Motor Operated Valve RRRMS

- Radiation Release Rate Monitoring System RMS Radiation Monitoring System RPS Reactor Protection System RPV Reactor Pressure Vessel RSCS Rod Sequence Control System RTD Resistance Temperature Detector RWCU Reactor Water Cleanup System RWM Rod Worth Minimizer SA Service Air SAR Safety Analysis Report SCN Standard Change Notice SDB Short Delay Band SGTS Standby Gas Treatment System SIL Section Instruction Letter SOS Shift Operations Supervisor SPDS Safety Parameter Display System SRM Source Range Monitor ST Special Test STPU Short Time Pick-up TACF Temporary Alteration Control Form TCV Temperature Control Valve TDR Time Delay Relay TIP Traversing Incore Probe TOL Thermal Overload TS Technical Specifications UF Under Frequency UFSAR Updated Final Safety Analysis Report UNID Unit Identification UV Under Voltage WO Work Order WP Work Plan

0 SAFETY EVALUATIONS FOR FINAL SAFETY ANALYSIS REPORT CHANGES

UFSAR 1.1 Project Identification, 2.3 Meteorology, and 2.4 Hydrology, Water Quality, and Marine Biology Units 1, 2 and 3 Chapter 1.1 (Project Identification)

The following sentence has been deleted from Section l.l.l.l (Applicant): "As of March 1982, TVA has 32,163 megawatts of generating capacity in service, and 14,422 megawatts (all of which are nuclear) under construction or in a deferred status."

This statement of fact has no direct impact on BFN. Because i,t contains no information which pertains to BFN directly and would have to be reverified regularly to be kept up to date, it is being deleted.

Chapter 2.3 (Meteorology)

In Section 2.3.4.2 (Wind Direction Persistence) the highest persistence duration at the 33-foot level for southeast winds was changed to 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The highest persistence durations at the 300-foot level for south-southeast and southeast winds were revised to 38 and 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months respectively.

In Section 2.3.5.2 (Precipitation), Reference 6 has been added to document the source for the maximum 1-hour rainfall for a 100-year frequency. Also, the sentence describing the information in Table 2.3-47 now includes the phrase

"...and normals for 1941 1970..." indicating the timeframe of some of the information in the table. The actual figures in the table have not changed.

In Section 2.3.5.3 (Snowfall), the reference for the maximum 24-hour snowfall was changed from'8 to 7. The document referenced has not changed.

In Section 2.3.6.2 (Tornadoes), the references have been renumbered because existing references 7 and 12 were deleted.

In the references on page 2.3-15, reference 5 was revised to correct its title. Former Reference 7, "Courtland Army Air Force Base, Alabama..." was deleted. It should have been removed in the 1981 update when onsite data for 1977-1979 was used. References 8-11 were renumbered 7-10 respectively.

Former Reference 12, "Storm Data, January 1972 through December 1976..." was deleted because it had no relevance to the text or to any tables of the

.UFSAR. References,13-15 were renumbered ll-'13 respectively.

The definition for stability class A in Tables 2.3-3 and 2.2-23 was revise'd to be "-1.9" rather than "1.9" as it was previously. The minus sign was inadvertently omitted when these tables were .added to the UFSAR. This is defined in NRC Regulatory Guide 1.23.

4

'k

V Minor editorial changes were made to Table 2.3-45.

Table 2.3-47 was revised to indicate the origin of data in each column (1968 1980 or 1941 1970). Also, the total days for precipitation of 0.01 inch or more (column 1) was revised from "118" to "119" and the normal annual precipitation (column 2) was revised from "52.4" to "52.2."

Chapter 2.4 (Hydrology, Water Quality, and Marine Biology)

In Section 2.4.4 (Water Use), the phrase "...not including BFN..." has been added. This is to indicate that the amount of surface water withdrawn from the Tennessee River by BFN is not listed in Table 2.4.4.

In Section 2.4.4.1 (Industrial), the phrase "...not including Browns Ferry..."

has been added. This is to clarify that the 172 million gallons of water withdrawn daily from Wheeler Reservoir does not include BPZ.

In Section 2.4.4.2 (Public), the phrase "...in accordance with the Radiological Monitoring Plan, Browns Ferry Nuclear Plant..." has been added.

This statement adds detail as to the source of monitoring requirements.

These administrative changes to UFSAR Chapters 1.1, 2.3, and 2.4 had no affect on any design basis accident in UFSAR Chapter 14. The changes provided more accurate and up-to-date information in the UFSAR. No new system or equipment interactions are introduced by this activity. No unreviewed safety questions were created and no Technical Specification changes were required.

UFSAR 2.2' Site Description Units 1, 2 and 3 t t v chapter was revised to reflect updated information on Population (Section 3'his 2.2.2) and Land Use (Section 2.2.3). Existing Table 2.2.1 (County Population Data) and Figures 2.2-5 through 2.2-16 (Population Distribution within 10 and 60 miles of the BFN Site 1970-2020) were deleted because. they are redundant to other tables and because they are not required by Regulatory Guide 1.70.

The following tables were revised to incorporate new data (revised titles given where applicable):

Table 2.2-2 1970 and 1980 Population of all Incorporated Places in 1980 Within 60-Mile Radius of Browns Ferry Site Tab'le 2.2-3 Cities and Towns Within 60 Miles Having 2,500 or More Residents in 1986

Table 2.2-4 Population Distribution Within 10 Miles of the Site Table 2.2-5 Population Distribution Within '60 Miles'f Site/60-Mile Population Distribution Table 2.2-6 Population Density at Various Distances from the Site Table 2.2-7 Peak Hour Recreation I

Visitation Within 10 Miles of the Site Table 2.2-8 Listing of 1989 School Enrollments and Industrial Employment Within 10 Miles of the Site Table 2.2-9 Browns Ferry Nuclear Plant Statistical Data for Nearby Counties The following Tables are being added:

Table 2.2-10 Hazardous River Traffic that Passes Browns Ferry Nuclear Plant:

1977-1987 (TONS)

Table 2.2-11 Waterborne Hazardous Material Traffic Survey Results 1988-89 The design basis accidents and anticipated operational transients are summarized in UFSAR Chapter 14.0 t t The information revised by this change to UFSAR Chapter 2.2 involved population distribution and land use of the areas surrounding the BFN site. None of the design basis events were impacted by the information in Chapter 2.2. No unreviewed safety questions were created and no Technical Specification changes resulted.

UFSAR 2.3.7.2 Data Acquisition System Units 1, 2 and 3 This change involved the description of the onsite meteorological measurement system. Data collected by this system is used in calculations of radiation doses from routine or accidental releases or radioactive material to the atmosphere. The uses include determining setpoints to limit doses from releases during normal operation (Appendix E) and calculating doses for control room habitability and design basis accidents (Sections 14.5.8 and 14.6, respectively). The change did not alter the meteorological measurement system or associated procedures. Also, this system is not listed as a safety system. Technical Specification 3.2.I/4.2.I'pecifies routine checks of required meteorological monitoring instrumentation channels to the plant control room to demonstrate that they are operable. The proposed change does not affect this Technical Specification because no change to the meteorological measurement system or data communication channels was involved. Technical Specification 3.8.B.1 specifies dose rate limits for routine operation. The release rates from components of the Gaseous Waste 0

Processing System are partly dependent on meteorological data used to calculate doses from releases, as indicated in Technical Specification 4.3.K.1 and UFSAR Appendix E. The text change did not change the meteorological data provided to the plant control room or any associated procedures and instrumentation. The meteorological measurement system and the data collection methods are unaltered by the change. No unreviewed safety question was created and no Technical Specification change resulted.

UFSAR 2.4 Hydrology, Water Quality, and Marine Biology Units 1, 2 and 3 This safety evaluation did not address the editorial changes. However the following four items were addressed:

1. Page 2.4-3 Changing the trapped volume of water from 100 x 106 cubic feet of water to 69.6 x 106 cubic feet of water.

2. Page 2.4-3 Added the following insert to the fifth paragraph: "The largest diffuser pipe reaches almost across the original river channel and extends above elevation 529 for its full length. Thus, the trapped pool following a postulated failure of Wheeler Dam is essentially divided into two parts with about 33 percent downstream and 67 percent upstream of the diffusers."

3. Page 2.4-4 Added the following insert to the first full paragraph:

"This is consistent with findings from TVA's system wide silt survey studies. TVA discontinued its silt survey program in 1965 because the earlier surveys show that the silting rate is small and is not a problem. TVA is presently reviewing its program needs for updating silt surveys on a system wide basis. Decisions about future updates for the cross sections around BFN and the frequency of the updates will be based upon this review."

4. Page 2.4-4 Removed the following: "...therefore, starting in 1971 rechecks of the silt range at representative locations will be made at about 10-year intervals. The results of these checks will be used to determine the need for future extensive surveys."

The downstream failure of Wheeler Dam cannot initiate any design basis accident as per Chapter 14 of the UFSAR.

Subsequent to having performed a new silt range survey the trapped volume of water in the pool after a Wheeler Dam failure was calculated to be 69.6 X 106 cubic feet as opposed to the original volume of 100 X 106 cubic feet. This pool volume is determined not to be critical for the shutdown cooling requirements of the plant. There is adequate minimum flow from tributaries and leakage from the Guntersville Dam to supply the required shutdown cooling volume of 36000 gpm or 80 cubic feet per second.

Due to the decrease in pool volume there is also a small change of the pool volume upstream and downstream of the RHRSW diffuser pipes which stretch almost across the original river channel at elevation 529. This small change has no impact on the shutdown cooling requirements of the plant.

TVA has performed silt surveys in the past and determined that the silt buildup is very slow. TVA is presently reviewing its program needs for updating silt surveys on a system wide basis. Decisions about future updates for the cross sections around BFN and the frequency of the updates will be based upon this review.

The suggested changes of the UFSAR did not increase the probability of an accident previously evaluated in Chapter 14 of the UFSAR. No unreviewed safety question was created and no Technical Specification change resulted.

UFSAR 3.4 Reactivity Control Mechanical Design Units 1, 2 and 3 UFSAR changes which affect the description of the CRD mechanism, CRD hydraulic system, and scram discharge volume subsystems of the CRD System have been identified as a result of the UFSAR verification effort. The changes which affect the CRD system are listed below.

Revise the distance to the active fuel zone as shown on Figure 3.4-4 to make the distance more general and revise the diameter of the guide tube.

2. Replace Figure 3.4-5 with a clearer, updated figure from GEK-94891, Attachment 2.

3~ Replace Figure 3.4-7 with an updated figure which simplifies the CRD flow information from GEK-32524A Figure 2-2 and TVA CRD flow diagrams. The system return line and TVA UNIDs were added to the figure.

4~ Delete the tensile strength and allowable load capacity of the CRD flange bolts in Section 3.4.6.4(c) to remove unnecessary detail.

5. Delete the tensile strength of the flange and housing in Section 3.4.6.4.(d) to remove unnecessary detail.

6. Delete the tensile strength of the CRD housing and the hoop stress in Section 3.4.6.4(e) to remove unnecessary detail.

7~ Revise the diameter of the guide tube and the velocity limiter and the width of the annulus between the limiter and the guide tube in Section 3.4.5.1.2.

The changes to Figures 3.4-4 and 3.4-5 are to simplify and update the information shown for the CRD mechanisms. The changes to Figure 3.4-7 reflect, in a simplified manner, the flow information shown in references 2 and 3. This did not affect the as-constructed configuration or operation of the CRD hydraulic system or scram discharge volume subsystems of the CRD system. The control rod drop accident is the only credible accident for the CRD system. No physical or operational alterations have been made. The proposed change removed details from the descriptions of malfunctions already considered in the UFSAR. These changes did not physically or operationally alter the system. The CRD velocity limiter limits the consequences of an uncoupled rod. The consequences of the CRD housing failure are limi.ted by the CRD housing support. The consequences of the hydraulic line break (which is a small break LOCA) are limited by the ECCS. These changes did not affect the velocity limiters, the CRD housing supports or the ECCS. For the CRD system malfunctions, UFSAR Section 3.4.6.4,assumes that all the drive flange bolts fail, that the flange and attached drive detach from the vessel and that the rupture in the drive housing has a flow area equivalent to the flow in the annular area in the drive. The proposed changes did not alter any of these assumptions or the results of the evaluations. No unreviewed safety question was created and no Technical Specification change resulted.

UFSAR 5.3 Secondary Containment System Units 1, 2 and 3 t v This change revised the description of Reactor Building penetrations in the Section 5.3.3.5, page 5.3-11, second paragraph, to:

1. Reflect that one spare mechanical penetration and several small instrumentation and control penetrations are exceptions to the UFSAR criteria that require through wall penetrations to exit the Reactor Building "...underground or enter into an adjoining concrete structure..."

2. Reword the description "underground" to read "below grade level" to more accurately reflect the actual plant configuration (i.e., not all these penetrations are buried under dirt).

The secondary containment serves a passive safety function during DBAs by maintaining its relative leak tightness to prevent radioactive releases to the environment. The through-wall secondary containment penetrations were located, in the majority of cases, such that this function could be maintained during a tornado event. Since there is no high speed rotating equipment outside the Reactor Building near these penetrations, the only credible missile threat to these penetrations is from tornados. This change, therefore, affected only the tornado event which is not proposed to occur coincident with any DBA so no DBAs are affected by this change.

All failure modes associated with this change affect only events which have been previously evaluated, i.e., breach of the secondary containment and safety equipment damage from an external missile. The increase in the probability of occurrence of both of these failure modes has been shown to be insignificant, so no new type of credible accident can be postulated to occur

, as a result of this change. No unreviewed safety question was created and no Technical Specification change resulted.

0 UFSAR Table 6.5-1 Operational Sequence of Core Standby Cooling Systems Units 1, 2 and 3 SEBFCAQ 900091, RO evaluates the safety impact of extended DG ready to load times for BFN. This change was incorporated into Section 6.5 and Table 6.5-1. The DGs committed to start and be ready to load within 13 seconds of an initiating signal. An engineering evaluation has determined that the DG loading sequences are such that the DGs can be started and be in a ready to load configuration in at least 13 seconds. However, the electrical breaker that connects the DG to the 4160 Volt emergency power distribution system requires an additional 4.5 seconds in order to close. Closing this breaker allows the DG to supply power to the 4160 Volt emergency power distribution system. This 4.5 second extension has been evaluated by GE and found not to adversely affect the peak cladding temperature following a limiting design basis accident coupled with a single failure.

The text and Table 6.5-1 were revised to show the extended DG ready to load time (13 + 4.5 = 17.5 seconds) and new corresponding loading sequence.

The change to the UFSAR did not adversely affect any safety related equipment or function. This change, will only document the DG extended ready to load time which had been evaluated and shown not to change any previous

.evaluations. No plant modifications will be performed nor safety related functions affected as a result of this revision.

The DG can be operated under the most unfavorable anticipated condition without cooling water from the EECW system. This evaluation enveloped the DG start sequence as described on Table 6.5-1. Thus, there is no adverse affect on the DGs from the delay of EECM pumps to start. This modification to the text did not result in a decrease in equipment reliability or operability. No Technical Specification change or unreviewed safety question resulted.

UFSAR 7.0 Instrument Specification Tables Units 1, 2 and 3 CRLD BFEP-SQE-90266-ROO revised the instrument specification tables in UFSAR Sections 7.4 (Core Standby Cooling Control and Instrumentation), 7.7 (Reactor Manual Control System), and 7.8 (Reactor Vessel Instrumentation), by deleting the range and accuracy specifications for various system functions, provided appropriate instrument device numbers for the specific instrument described, and revised the trip setting column heading to include the analytical limits for instrumentation providing a safety related function. For those Tables (7.7-1 and 7.8-1) describing non-safety related instruments, the trip setting volume headings will remain in their present form. No changes or additions were proposed for the specified trip settings.

0 The instrumentation identified in the referenced tables provide control and/or alarm functions for both safety and non-safety functions and are typically necessary for the mitigation of the consequences associated with DBAs. The subject CRLD removed accuracy and range information which may or may not have been of any significance in the NRC review process. However, TVA committed to NRC to perform, within the DBVP, a series of calculations to provide an engineering basis for required ranges, accuracies, setpoints and" analytical limits, where appropriate.

Since identification of the range and accuracy information currently in the UFSAR was not based on engineering calculations, it is expected that some or all values will change. Although required ranges and accuracies may become less restrictive than currently specified, they are based on engineering calculations performed under an NRC approved program subject to NRC audit.

Therefore, the information currently provided in the UFSAR is unnecessary and was removed. No unreviewed safety question was created an no Technical Specification was involved.

UFSAR Table 7.2-1 Reactor Protection System Instrumentation Specifications Units 1, 2 and 3 t

lower analytical limit for low pressure in the scram air header from t'evised 50 PSIG to 45 PSIG. The Technical Specification for-this setpoint is to remain unchanged at 50 PSIG (Technical Specification Table 3.1.A). Purpose is to increase the margin between the current Technical Specification allowable value and the analytical limit value that is justifiable from the safety analysis. No change was made to any equipment. Only an analytical assumption was relaxed. Therefore, no real change to any event will occur. NRC requirement for 10 PSI margin between the setpoint and the scram outlet valve opening pressure is interpreted to apply to the Technical Specification allowable value (which remains at 50 PSIG and has at least 10 PSI margin). In this way, the NRC requirement is consistent with measured compliance surveillance testing. The 10 PSI range is required precisely for the reason that actual performance may be outside the Technical Specification the concept of analytical vs allowable values. Assuring that this uncertainty is no more that five PSI (50-45) is in compliance with NRC Bulletin which allow 10 PSI for this uncertainty. The acceptability of the setpoint is obviously dependent upon consistency of. TVA's testing criteria. They must be maintained consistent with the Technical Specification (y 50 PSIG for the pressure instruments) and maintenance testing of the 'opening pressure ((40 PSIG per GE 9582A and SIL-373). Based on this evaluation, there was no increase in the probability of an accident (incomplete scram or any other) previously considered for BFN. No unreviewed safety question was created and no Technical Specification change resulted.

1 I UFSAR Tables 7.2-1, 7.3-2, and 7.4-1 through -4 NSSS Instrument Analytical Limit Revisions Units 1, 2 and 3 The change revised the UFSAR per CRLD BFEP-EEB-90029 to incorporate revised analytical limits for main steam high flow, reactor water level, and nuclear system high pressure instruments.- Calculation ND-$0999-900038 issued a safety evaluation performed by GE and documented a change to the analytical limits for the following instruments which are listed in Tables 7.3-2, 7.4-1, and 7.4-4:

2-PdIS-001-013A-D 2-PdIS-001-025A-D 2-PdIS-001-036A-D 2-PdIS-001-050A-D 2-LIS-003-208B% D 2-'LIS-003-058A-D 2-LIS-003-052 2-LIS-003-062A Calculation ND-f0003-890063 issued a safety evaluation performed by GE and documented a change to the analytical limits for the following instruments which are listed in Tables 7.2-1, 7.4-2, 7.4-3, and 7.4-4:

2-LS-003-058A-D 2-PT-003-022AA, -022BB, -022C, -022D 2-LT-003-0203A-D, -184, -185 The analytical limit changes have the potential for delaying the signals which initiate safety functions. Analysis of analytical limits performed by GE demonstrates that the instruments will function adequately to assure proper initiation of their safety functions with the revised analytical limits considered. In all cases the increased response time was negligible. No unreviewed safety question was created and no Technical Specification change resulted.

UFSAR Tables 7.2-1, 7.3-2, and 7.4-4 Drywell High Pressure S~itches Units li 2 and 3 GE performed a calculation which documented a change to the upper analytical limit from 2.5 PSIG to 2.6 PSIG for the following Drywell High Pressure Switches:

2-PIS-064-056 A-D 2-PIS-064-057 A-D 2-PIS-064-058 A-H 2-PS-064-058 A-D In addition, a subsequent change to the lower analytical limit from 1.0 PSIG to 0.35 PSIG for PIS-64-058E-H was received. The setpoints and the allowable value for the subject instruments will not be changed. Therefore, there will be an increase in the margin between the current Technical Specification allowable value and the analytical limits that were justified by the GE analysis or by TVA calculations.

Additionally, since the instrument setpoints for the high drywell pressure signals are not being revised, there is no change in the ability of the affected systems to operate for mitigation of design basis events. The key events where high drywell pressure signals are needed are the loss of coolant accidents. For the bounding cases, no credit is taken for the scram signal from high drywell pressure since the worst case starts at the low level scram. For milder cases, the slight delay of scram from high drywell pressure which could occur due to the increased analytical limit will not increase the severity of the event. For all other actions derived from this signal, the small delay possible with a higher analytical limit is negligible relative to the action tie taken (e.g., isolations take several seconds, ECCS injections takes many seconds). This evaluation is principally taken from GE analysis of analytical limits. The drywell containment interlock has an upper and a lower analytical limit which were chosen to assure that containment spray can be initiated before the drywell pressure exceeds 2.6 PSIG and prevents containment spray at a pressure less than 0.35 PSIG. Initiating containment spray at less than 0.35 PSIG could damage the drywell because a decrease in pressure below atmospheric could occur. No unreviewed safety question was created and no Technical Specification change was required.

UFSAR 7.12 Process Radiation Monitoring System and 10.6, Reactor Building Closed Cooling Water System Units 1, 2 and 3 t v The changes to the UFSAR involved the 'Radiation Monitoring system and the RBCCW system. There were no changes directly related to the RMS other than a statement in the UFSAR Section related to the source of makeup water for the RBCCW system. The UFSAR changes eliminated certain water quality parameters on the RBCCW cooling water. No equipment was added, modified, or removed from BFN as a result of the UFSAR changes. The text changes did not result in any new or different design input. However, the text changes provided clarifications and made the UFSAR consistent with other design documents and surveillance instructions and did not introduce any new failure modes or alter existing failure modes. No unreviewed safety question was created and no Technical Specification change resulted.

\

UFSAR 7.13-2 and 7.13.4 Area Radiation Monitoring Units 1, 2 and 3 The subject CRLD revised the UFSAR to delete incorrect and unnecessary information and to correct the identified elevation location of specific radiation monitor detectors. Additionally, some editorial changes are made.

The ARM system is a non-safety related system and has no safety function to perform. -A review of the BFN-Unit 2 'Technical Specifications indicated that the ARM system is not addressed in the Technical Specifications nor are any individual ARM system components including the portable ARM detector calibration unit. Therefore, the equipment identified is not associated in any way with a Technical Specification Margin of Safety and, thus, there was no impact on any Margin of Safety for these UFSAR changes. Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the UFSAR was increased or created.

Therefore, no unreviewed safety question was created and no Technical Specification change was required.

UFSAR 8.5 Standby AC Power Supply and Distribution Units 1, 2 and 3 t v t The subject CRLD revised the UFSAR to agree with the actual configuration for the DG starting circuits by changing the statement on the operation of the mode switch bypass to "only on fast start". The diesel generators and their operational mode selector switches are associated with this text change because, the current plant configuration does not bypass the mode selector switch on all diesel generator starts as stated in the UFSAR. There are no credible failure modes associated with this text change which have not already been analyzed in Sections 8.5.3.2 and 8.5.4.4. The change to the UFSAR described the current configuration of the diesel generator operation under normal and emergency conditions. Failure of a diesel generator during an abnormal transient or design basis accident remains bounded by single failure analysis. The existing configuration does not alter the function of the diesel generator. No unreviewed safety question was created and no Technical Specification changes resulted.

UFSAR 8.5.3 125V DC Diesel Control Power Units 1, 2 and 3 As stated in UFSAR Section 8.5.3.2, each diesel has its own battery with a normal and alternate battery charger. Each battery consists of 60 cells (20 three-cell containers) with a nominal terminal voltage of 125V DC at 77 degrees F and a minimum terminal voltage of 105V DC. The stated discharge rate for each battery is 76 amps for 30 minutes and 148 amps for 1 minute.

The diesel generator battery capacity calculation showed that the worst case calculated load profiles are enveloped by vendor specification requirements.

amp'20-There are no systems affected by this change. The only parameter affected is the change in the continuous load current requirement from 4.2 amps to 5.0

0 The small increase in continuous load requirement was presumed to be due primarily to equipment change-outs using qualified substitutions for replacement relays and/or timers. Since the battery is capable of discharging at a rate to 76 amps for 30 minutes and the worst case loading conditions (which include this slightly higher continuous load) determined by calculation are easily bounded by that capability, there was no impact on the as described in the UFSAR. The continuous battery load of 5 amps batteries'apabilities is for control power loads. Control components have been verified by walk-down- and were also found to be of identical or similar make, construction, and electrical load. Therefore, there are no credible failure modes associated with this slightly higher continuous electrical load. No unreviewed safety question was created and no Technical Specification change resulted.

UFSAR 8.6.3 250V DC Power Supply and Distribution Description Units 1, 2 and 3 t v t'he 250V DC power system provides control power and motive power for containment isolation and core standby cooling systems and components under normal, transient and emergency conditions. This change deleted reference to "106 percent" of rated value for the specified battery charger output voltage regulation accuracy. There is no such limitation on battery charger regulation accuracy. The voltage regulation accuracy of +/- 0.5 percent as stated is valid over the entire range of voltages and is not limited as implied by this number. The battery charger design and configuration were not altered in any way by this 'text change. The reference to a limitation on design voltage output regulation accuracy at 106 percent of rated voltage was meaningless and potentially misleading. The battery charger does not have one rated output voltage. It is rated over a range of voltages between 240 VDC and 295 VDC. Voltage regulation accuracy of +/- 0.5 percent remains constant on the steady state output of the battery chargers throughout this range. No unreviewed safety question was created and no Technical Specification change resulted.

UFSAR 8.6.4 250V DC Power Supply and Distribution Safety Evaluation Units 1 ~ 2 and 3 The 250V DC Power System is safety-related and provides a highly reliable source of control and motive power as required for engineered safeguard systems so that no single, credible event can disable the containment isolation and core standby cooling functions and their supporting control power sources and circuits. The 250V DC plant batteries are required to be capable of supplying their connected loads for 30 minutes (Section 8.6.2).

Calculation ED-f2000-87041 revision 1 verifies that the batteries are capable of performing their required function. The existing UFSAR refers to Figures 8.6-4a, b, c, and d to confirm this. Figures 8.6-4a, b, c, and d do

not reflect the existing 250V DC load requirements and were replaced by Calculation ED-Q2000-87041 revision l. The proposed change removed the UFSAR Figures and all references to them. The batteries continue to be capable of supplying their required loads for the required 30 minutes and meet all other commitments. The effect of these changes was'o remove incorrect and out of date information from the UFSAR. There is no effect on battery or 250 Volt DC Power System design capabilities or functional performance. These text changes only removed incorrect and out of date information from the UFSAR.

The result of these change was to ensure only accurate, verified design information is provided in the UFSAR. Based on this and a review of UFSAR Section 8.6, no new failure modes for this system are postulated.

Since the proposed text changes did not alter in any way the original design, there is no reduction in accident mitigation capability. UFSAR Section 8.6.2 requires that the batteries must be capable of supplying their connected loads for 30 minutes. Battery capability to meet this requirement was confirmed by Calculation ED-Q2000-87041 revision 1. Because there was no impact on system or equipment design, configuration or functional performance, the 250V DC Power system functional capability was not affected and there was no affect on any Technical Specification margin of safety resulting from the subject UFSAR changes. No unreviewed safety question was created and no Technical Specification change resulted.

UFSAR 8.7.3.2 Plant Preferred 240/120V AC System Units 1, 2 and 3 The proposed change was to revise UFSAR Section 8.7.3.2 per CRLD BFEP-EEB-89004 to reflect that the plant preferred power system bus transfer from the alternate power supply back to the normal power supply is automatic rather than manual (as shown on UFSAR Figure 8.7-2). The power system and its loads are non-safety-related'nd serve no direct accident mitigation function. Further, the effects of a transient initiator would tend to be reduced owing to the automatic feature of the bus transfer. The use of an automatic transfer back to normal power would not result in any perturbations to the plant operation except for loading the lighting panel again since the loads would be on the MG set already. No new accident types are postulated as a result. No unreviewed safety question was created and no Technical Specification change resulted.

UFSAR 8.8.2 ,24V DC Power System Units 1, 2 and 3 D t' The 24V DC Power System is non-safety-related and provides 24V DC power to SRNs and IRNs and their trip auxiliaries, trip auxiliaries for off 'gas radiation monitor and timer control, and radiation monitors for stack gas, linear off gas, RHR Service Water effluent, RBCCW and RCW effluent during all modes of plant operation. The 24V DC batteries are required to be capable of supplying their connected load for three hours (SAR Section 8.8.2.2).

Calculation'22-

0 ED-Q2000-87049 revision 1 verifies that the batteries are capable of performing their required function. The existing UFSAR text (Section 8.8.2.3) also states that these batteries are capable of supplying their connected loads for up to approximately 3 1/2 hours and that this is supported by Table 8.8-2. Based on a review of Calculation ED-Q2000-87049 revision 1, the information in Table 8.8-2's not correct and there is no justification for claiming a 3 1/2 hour capability for these batteries. The UFSAR change removed referenceto the table and all reference to a 3 1/2 hour capability and deleted Table 8.8-2. The batteries continue to be capable of supplying their connected loads for the required three hours and meet all other regulatory commitments. The effect of these changes was to remove incorrect and unverifiable information from the UFSAR. There was no 'effect on the battery or 24V DC Power System design capabili.ties, function, or performance.

No unreviewed safety question was created and no Technical Specification change resulted.

UFSAR 8.9.2.1 Raceways General Plant Units 1, '2 and 3 t n t v t' revision to the UFSAR was required to reflect the current voltage level separation requirements for safety-related and non-safety-related cabling.

Additionally, a clarification of the voltage level separation requirements for all cables installed prior to the initial issue of General Design Criteria BFN-50-758 RO was required. The subject revisions to the UFSAR implemented the corrective action for CAQR BFP880206.

CRLD-BFEP-SWE-90069 Rl revised UFSAR Section 8.9. 2.1 to resolve discrepancies in cable tray voltage level designations including a more concise definition of voltage level applications, and cable routing instructions initiated after July 31, 1987. The subject CRLD also revised UFSAR Section 8.9.6 to better define placement of cables in cable trays and to recognize additional cable routing instructions before and after the July 31, 1987.

The subject UFSAR revisions do not alter the original licensing basis except that the routing of higher voltage cables with lower voltage cables is allowed on a case-by-case basis by General Design Criteria BFN-50-758. Since each case will be reviewed by 1VA'Engineering against BFN design criteria which includes BFN-50-729, "Single Failure Criteria for Fluid and Electrical Safety-Related Systems," and the installation verified acceptable by conservative calculational methods, BFN will maintain the single failure criteria specified by Title 10 CFR in its design configuration. The clarifications defined by the UFSAR revisions do not change the function or operation of safety-related equipment and systems. Since the UFSAR system descriptions remain unchanged and design criteria compliance is mandatory prior to startup (e.g., single failure criteria), the analyses in the UFSAR envelope the possible cable failure types and no new, different types of cable failure are postulated to occur as a result of adhering to the requirements specified in the UFSAR text and table revisions. No unreviewed safety question was created and no Technical Specification change resulted.

UFSAR 9.3 Solid Radwaste System Units 1, 2 and 3 SAR changes which affect the description of the quantities and radioactivity levels of solid radioactive wastes, spent resin tank construction, storage period, and personnel exposure have been identified as a result of the UFSAR verification effort. These changes revise the UFSAR to remove details on quantities of solid radwastes and sludges, activity levels, personnel exposure rates, and storage time. These details were used to estimate the throughput of the radwaste system and the resulting personnel exposure rate. The throughput is monitored and reported to NRC in accordance with the Technical Specifications. Personnel exposure rates are monitored by RADCON so that the limits of 10 CFR 20 are not exceeded. Removing these values did not affect the safe operation of BFN. These UFSAR changes did not directly affect the equipment or operation of the solid radwaste system. The actual programs for solidifying and packaging wastes were not affected; thus, the requirements of 10 CFR 20, 10 CFR 71, and the burial site were not violated by these UFSAR changes. No unreviewed safety question was created and no Technical Specification change resulted.

FSAR 10.6 Reactor Building Closed Cooling Water System Units 1, 2 and 3 t v The UFSAR currently states that the RBCCW system is designed to Class I specifications. The UFSAR change (CRLD Number BFEP-NN-90004 Revision 1) reclassified the portion of RBCCW outside the primary containment isolation boundary to Class II. Section 10.6 and Appendix F of the UFSAR was revised to reflect this reclassification. Responses to AEC questions 14.1, C.l and D.3 of Narch 15, 1971 which were included in volume 7 of Amendment 70 of the FSAR were revised accordingly. This revision corrected several typographical errors identified during a review of the associated UFSAR change package and superseded SABFNSAR90012 Revision 0 in its entirety. The objective of Phase I of the DBVP is in part, to evaluate the systems and portions of systems required for safe shutdown to ensure that their configuration satisfies the design basis and is in conformance with BFN licensing commitments. A review of the documents generated for the DBVP (Restart Design Criteria, Safe Shutdown Analysis, and the System Evaluation Report) indicated that the portion of RBCCW affected by this change was not considered to be required for safe shutdown. Therefore, reclassifying RBCCW outside the primary containment isolation boundary to Class II did not impact any of the evaluations performed as part of the DBVP in establishing the design basis documents. This change introduced an increased probability of a seismic induced failure of the RBCCW piping outside the primary containment isolation boundary thus preventing RBCCW from supplying cooling water to the equipment it serves. Failure of the equipment and systems serviced by RBCCW to function is considered in the UFSAR. No unreviewed safety question has been created and no Technical Specification change was required.

UFSAR 10.10 EECW System Units 1, 2 and 3

~

t ~

Drawings 1-47E859-1 R023 and 3-47E859-2 R008 show the required pressure and temperature conditions for the EECW system. These drawings are UFSAR Figures 10.10-1a and 10.10-lb respectively. References to design pressure and temperature conditions for specific portions of the system will be removed from the text of UFSAR Section 10.10. Reference will be made to these Figures only. Since the drawings depict the current conditions in the plant, there will be no field work required. However, the changes to the design conditions axe a change from the conditions stated in UFSAR Section 10.10. A calculation revision documented that the pressure/temperature ratings of the piping, valves and components are acceptable for the expected operational conditions including 15 percent overpressure 10 percent of the operating period and 20 percent overpressure 1 percent of the operating period. The basis for revision of the system design parameters was included in calculation MD-Q2067-880344 revision 5. The calculation provided the design input basis for pressure/temperature ratings of specific portions of the EECW System as shown on drawings 1-47E859-1 R023 and 3-47E859-2 R008. Since no changes were made to the plant, this activity did not increase the consequences, of a malfunction of equipment important to safety. No unreviewed safety question was required and no Technical Specification change was created.

UFSAR 10.11.5 Fire Protection Systems Analysis Units 1, 2 and 3 The change revised UFSAR Section 10.11.5 to correctly indicate the flood level switches on elevation 519 are two inches off the floor. The section was also revised to correctly describe the locations of the switches. The six flood level switches on reactor building elevation 519 each independently actuate an alarm (LA-77-25A through 25F) on control room panel 9-4. When one or more of the alarms actuate, ARP-9-4 requires the operator to dispatch an assistant unit operator to visually check the location of the level switch(es) for flooding. Mitigation of flooding conditions on elevation 519 is desirable before the water level reaches two feet (i.e., elevation 521) and potentially floods out the RHR, HPCI, RCIC, and core spray pumps. Detection of a flooding condition at two inches, rather than six inches as previously stated, would allow more time to determine the cause of the flood and take mitigating actions. Hence, the two inch setting reduces the possibility of a flooding condition causing the safety-related pumps to malfunction. The level setting of a switch does not increase the probability of a malfunction of the switch itself. All QA and maintenance requirements for the switches and associated circuitry remain as before. No unreviewed safety question was created and no Technical Specification change resulted.

J" UFSAR D

10.20 Auxiliary Boiler Systems Units 1, 2 and 3 v

1, 2, and 3, respectively, installed t'CNs W0106A, W0107A, and W0108A for units blind flanges on the deaeration lines between the auxiliary boilers and the condenser hotwells. This prevents stress corrosion cracking in the condenser tubes by ammonia which is used to treat the boiler feedwater and has 1'caked past closed valves in the line to the condenser hotwell. CRLD BFEP-MN-90078 RO revises UFSAR Section 10.20 to show that blind flanges were added to the deaeration line to the condenser hotwell. DCNs W0106A, W0107A, and W0108A removed piping and added blind flanges in the portion of the deaeration lines

'hat are normally isolated from the condenser hotwell by two closed valves (one is locked closed) and isolated from the boiler by one closed valve. This line is not Seismic Class I and can be assumed to break. The Auxiliary Boiler System is associated with AOTs and accidents that could result in a release of radioactive material to the environment to such an extent that the limits of 10 CFR 20 are exceeded. Manual valves in the Auxiliary Boiler System are required to remain closed to maintain primary and secondary containment isolation during plant operation. The Auxiliary Boiler System does not perform any function which prevents any AOTs or accidents. No unreviewed safety question was created or Technical Specification change resulted.

UFSAR 12.2.2.6 Sacrificial Shield Wall Units 1, 2 and 3 The sacrificial shield wall provides a biological shield for protection of personnel from gamma radiation from activated vessel components, a neutron shield to prevent activation of the drywell components during operation, and a means of supporting the drywell pipe hangers and access platform.

Due to the reanalysis of the unit 2 elevation 584 foot 9 1/2 inch drywell platform, the concrete strength must be considered at the sacrificial shield wall platform anchorages. The design approach utilized the steel beams between the sacrificial shield wall structural columns as reinforcement in the concrete. Also due to these beams being embedded, their end attachments were assumed to be fixed. The maximum height of concrete needed for structural purposes in this affected area is the lower 10.5 feet. Per the original design and as noted on the design drawing, the lower 19 feet 6 3/4 inch of the sacrificial shield wall concrete strength is 4000 psi.

The concrete strength of the, Sacrificial Shield Wall to 2.5 feet above the 8.0 feet, as stated in the UFSAR, was evaluated in the radiation intensities calculation and the structural analysis which demonstrated that the radiation dosages at the elevation 584 feet 9 1/2 inch drywell platform anchorages have no significant effect on the concrete strength of the Sacrificial Shield Wall and that the concrete strength is adequate at that elevation. This change of lower 10.5 feet of the wall from previous lower eight feet of the wall was added to UFSAR Section 12.2.2.6. The concrete strength as specified on drawing 48N445 meets the applicable code requirements for the lower 10.5 feet. No unreviewed safety questions were created and no Technical Specification change resulted.

0 UFSAR 12.2.4 Reinforced Concrete Chimney Units 1, 2 and 3 UFSAR change request CRDL-BFEP-SWC-90033, ROO identified proposed changes to the BFN UFSAR Section 12.2.4 Reinforced Concrete Chimney.

The following changes concern the UFSAR description of the required design cases and"allowable stresses for the chimney shell. The subject CRLD proposed to correct Table 12.2-25 consistent with the design criteria and the ACI Code determination of concrete and steel allowable stresses, referenced the applicable ACI Code, corrected an incorrectly identified allowable for Design Case 3 for consistency with the actual calculated allowable, and added a reference to the ACI code for Design Case 1. Added "DL +" to Design Cases 1, 2, and 3, and revised the concrete and steel allowable stresses. In addition, a reference was added to ACI 307-69 as follows: "ABACI 307-69."

These changes were made to reflect the design conditions stated in Design Criteria BFN-SO-C-7100, Revision 1, Section 5.1, Table 4-1; Attachment C, Section 5.2.1; and Attachment C, Table 15-18. The design criteria, in fact, reflect the design basis ACI code reference A.2.7 requirements and did not initiate or describe any change in plant design or operating procedures.

Thus, no impact on any UFSAR evaluation resulted. No unreviewed safety question was created and no Technical Specification change resulted.

FSAR 13.2 Organization and Responsibility, and FSAR 13.7 Records Units 1, 2 and 3 v

Chapters 13.2 (Organization and Responsibility) and 13.7 (Records) were t'FSAR revised by this change. The information in Chapter 13.2 was completely rewritten to describe the current BFN and TVA Nuclear Power organizations.

The following changes were made to Chapter 13.7.

(1) A revision was made to indicate the "Site Records Management Supervisor" is now called the "Site Document Control and Records (DORM) Management Manager."

(2) A typographical error in paragraph 3 was corrected by changing "revision" to "reviewer."

(3) The term "Records Management Staff" was revised to "DORM" in paragraphs 3 and 5.

(4) In paragraph 6, "Plant Planning Support and the Records Management groups" was revised to "Work Control Support Group and retained by Site DORM." This is to clarify responsibilities and organization titles.

I (5) Paragraph 7 was rewritten to indicate that Radiological Control maintains the records and DCRM retains them.

Updating UFSAR Chapters 13.2 and 13.7 to reflect the current BFN and TVA, Nuclear Power organizational structures had no affect on any design basis accident in UFSAR Chapter 14. The change provided more accurate and up-to-date information in the UFSAR. No new system or equipment interactions were introduced by this activity. No unreviewed safety question or Technical Specification change resulted.

UFSAR 13.3 Training Program Units 1, 2 and 3 t' Ev The change involved revising the text of UFSAR Section 13.3 to update to current training program descriptions. The changes updated the UFSAR text to conform with 10 CFR 55, as revised May 25, 1987, with regards to obtaining and maintaining a Reactor Operator License or Senior Reactor Operator License; added text to the description of the non-licensed training program; added text to more accurately describe the General Employee Training Program; corrected organizational and personnel titles; updated the BFN Operator training to conform with STD 7.1.905, "Nuclear Operator Training Program"; and added program descriptions to more accurately describe current training. The proposed change to BFN Section 13.3 did not impact any BFN Technical Specifications nor adversely affect any BFN system, structure, component, function, or parameter, either directly or indirectly. Therefore, no acceptance limits in the basis for any BFN Technical Specification or UFSAR was exceeded by this change. As a result, no change to the Technical Specifications was required, nor was an unreviewed safety question introduced.'FSAR 13 6.4 Maintenance Instructions Units 1, 2 and 3 t' v The requir'ed change involved revising the text of UFSAR Section 13.6, Subsection 13.6.4, "Maintenance Instructions," to address the new Work Request/Work Order process. UFSAR Subsection 13.6.4 state'd, in part that maintenance is initiated through the maintenance request program and a preventive maintenance program. The change added Work Request/Work Order process to the text of this Subsection. There are no failure modes associated with this change'. The change did not affect any other information in any other section of the UFSAR and did not alter any initial conditions, assumptions or systems response time for any of the accidents or transients in BFN UFSAR Section 14, Plant Safety Analysis. Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the UFSAR was increased or created. Therefore, no unreviewed safety question was created and no Technical Specification changes were required.

UFSAR 13.10 Process Computer Units 1, 2 and 3 UFSAR Chapter 13.10 page 13.10-5 (Section for test number 13 level 2 acceptance criteria) specifies that the CLIP program will be used for comparing the process computer calculated MCPR. The CLIP program is an off-line thermal limits calculation program residing on the plant PRIME computer. CLIP was replaced with the PATH program that resides on an off-line VAX computer. This replacement of the CLIP program was driven by more stringent software qualification requirements and the increased operational flexibility of using the VAX (dedicated to engineering support) in lieu of the plant PRIME computer.

The revision changed the name "CLIP" in the UFSAR (two locations on page 13.10-5) to the more generic and descriptive term "qualified backup method."

Typically, the backup methods use the same algorithms as the primary methods on the process computer and qualification testing includes comparisons to actual process computer calculations. This is true of both the CLIP and PATH programs which use the same algorithms in the backup method.

The PATH program was qualified in ac'cordance with the MBI-02 procedure which in tu'rn implements the upper tier requirements found in PMP 1801.06 and NFP 8.0. Although these specific procedures may be replaced in the future, the basic requirements of adequate documentation and program verification will remain the same. The new methods are more stringently qualified than the original CLIP program. Therefore, no adverse impact on monitoring of these limits is expected. The software is used as a backup method for monitoring the thermal limits during operation. These limits in turn are based upon the pre-analyzed events. The change affected monitoring only and can neither directly or indirectly affect the type of events. Plant response to an event is unmodified by this change. A credible failure mode would be an inaccurate calculation during a period of process computer unavailability. This was the same credible failure mode associated w'ith the current backup method.

The proposed change did not affect plant components, structures, or systems.

The change allowed the use of a new backup system for core thermal limits monitoring that was developed and qualified under more stringent software requirements than those imposed upon the original backup software. Therefore, no unreviewed safety questions were created and no Technical Specification changes were required.

UFSAR 13.10.2.5 Core Power Distributions Units 1, 2 and 3 D

' n t v The two changes in the wording of UFSAR Chapter 13.10.2.5 were:

1. Remove the requirement to perform repetitive individual TIP traces to calculate TIP random noise (as well as the geometric component of the Total TIP uncertainty), and

2. Remove the requirement to adjust the MCPR operating limit if the criteria for the Total TIP uncertainty is not met.

The only potential effect was in regard to the anticipated operational transients. The MCPR operating limit is defined to limit operation within previously analyzed bounds for the limiting transient(s). Fuel failure could potentially occur if a limiting transient began while the reactor core was operating below the NCPR operating limit. The steady state MCPR operating limit is established to quantify the analytical bounds for the initiation of one (or more) of the anticipated operating transient events'he consequences of a MCPR limiting transient would be uncertain if the event initiated while the core was operating below the established MCPR operating limit. However, for an increase in the TIP uncertainty to place the core in this state, the calculated power for the lead bundle would have to be lower than actually exists.

The total TIP uncertainty is actually a measure of the symmetry of the TIP readings across the TIP symmetry line, and random variations in the instrument readings. High uncertainties reflect asymmetries in these readings.

A lower TIP reading occurring adjacent to the lead bundle would result in lower calculated powers. The MCPR calculated for this lead bundle would increase. However, due to core-side normalization of the TIP readings, the symmetric TIP string will show higher than actual readings. As a result of the higher TIP reading, the symmetric bundles will appear to have higher powers and a decrease in the calculated MCPR. The net result is a change in lead bundles with a conservative change in the lead bundle calculated powers and MCPR. The changes did not involve an unreviewed safety question. No changes were made to plant components, structures, or systems. The changes were made to simplify a plant instruction and remove what the reactor/fuel vendor (GE) considers to be an obsolete requirement. Removal of the requirement to modify the NCPR operating limit if the criteria of 9.0 percent total TIP uncertainty is not met is justified since the NCPR is conservatively penalized with increasing TIP uncertainty. No Technical Specification change resulted.

UFSAR 14.5.4.1 Pressure Regulator Failure Units 1, 2 and 3 t v t This change was required to allow consistency in the assumed MCFL setting for the UFSAR analysis of the PRFO event with the recommendations contained in GE SIL number 502 Revision 1 dated January 19, 1990.

GE SIL-502 recommended that the MCFL setting be no lower than 115 percent of rated steam flow in order to protect against fuel damage during a Slow Closure of Single Turbine Control Valve Event. However, the existing UFSAR description of the analysis of a PRFO Event containment in Section 14.5.4.1 describes an analysis for an MCFL setting of 115 percent as the worse case.

Since the PRFO event is limiting for high MCFL settings, the existing UFSAR text is inconsistent with a MCFL setting above 115 percent as recommended by GE SIL-502.

The maximum steam flow demand allowed is controlled by the MCFL setting which determines how far the turbine and bypass valves are allowed to open. The revision modified the UFSAR description of the PRFO event to describe an analysis for the maximum achievable steam flow demand of 125 percent. This change made the UFSAR event description consistent with the actual plant configuration but did not change the design in any substantial manner. No unreviewed safety question was created and no Technical Specification change resulted.

UFSAR Appendix M Report on Pipe Failure Outside Containment Units 1, 2 8Ild 3 v

One of the functions of the UFSAR Update was to ensure the adequacy of the documentation supporting individual UFSAR Change Packages. Additionally, TVA STD-6.1.3 and BPN SDSP 27.1 required that UFSAR changes be fully supported by 10 CFR 50.59 Safety Reviews unless exempted by procedure. CRLD-BFEP-CEB-90007 ROO identified proposed changes to the BFN UFSAR Appendix M, "Report on Pipe Failure Outside Containment in the BFNP." Therefore, a 10 CFR 50.59 Safety Review was required. UFSAR Appendix M was being revised to reflect where relief from the effects of postulated pipe failures was being incorporated into the BFN design basis consistent with industry accepted practice.

NUREG/CR-2913, "Two Phase Jet Loads," evaluated jet and target pressures for axisymmetric target geometries. Based on findings in this report, other utilities have utilized a position that jet impingement loading from steam or flashing subcooled liquid breaks is not significant on targets located at distances from the break greater than 10 times the inside diameter of the broken pipe. Among the utilities utilizing this position is South Texas Project, docket numbers STN 50-498, STN 50-499. BFN proposed to utilize the following position in the performance of jet impingement evaluation: "When the jet consists of steam or flashing subcooled liquid, unprotected equipment/components located at a distance greater than 10 diameters (broken pipe ID) from the break, or equivalent diameter from the crack, shall be assumed to be undamaged by the jet without further analyses, provided that the environmental qualification of the target is not exceeded." The incorporation of more realistic jet separation criteria for evaluating localized jet effects on plant modifications represented a screening criteria change only. No plant modifications were associated with this criteria change. In addition, pipe rupture events which are evaluated in the UFSAR address the global effects of the event such as subcompartment pressurization, plant response, etc., and are not dependent on techniques used to screen potential jet impingement interactions. Therefore, this position did not create a new unanalyzed type of malfunction nor did it increase the probability of an event not previously evaluated in the UFSAR. No unreviewed safety question was created or Technical Specification change resulted.

SAFETY EVALUATIONS FOR NEM INSTRUCTIONS 2-AOI-64-8 Primary Containment High Hydrogen and/or Oxygen Unit 2 This procedure provides the operator with directions for mitigating the effects of hydrogen and oxygen in the Primary Containment. The purpose of 2-AOI-64-8 is to identify the symptoms and the operator actions necessary to control combustible gases inside the Primary Containment. As such, it implements the manual actions assumed in the UFSAR following a Loss of Coolant Accident (see UFSAR Section 14.6.3) for control of combustible gases. This description is found in the UFSAR Section 5.2.6, Combustible Gas Control in the Primary Containment. 2-AOI-64-8 specifies the correct technical guidance, in the proper format, to accomplish combustible gas control in the Primary Containment. 2-AOI-64-8 was verified and validated in accordance with plant procedures, and also to the same standards of PMI-12.8 and PMI-12.9. No unreviewed safety question was involved and no changes to the Technical Specifications were required.

2-TI-147A Core Reload Unit 2 This new procedure will be used in place of the generic fuel loading procedure (TI-147) for loading the Unit 2 cycle 6 core. The major differences between the new procedure (2-TI-147A) and the generic one are listed below:

1. The new procedure references unit specific procedures where possible.

2. Current titles are used (e.g., "Nuclear Engineer" changed to "Reactor Engineer" ).

3. Redundant statements of requirements from 2-GOI-100-3 and 2-SI-4.10.B have been deleted unless they are something that the reactor engineer should be continuously aware.

4. Delete operations verifications of prerequisites which are required by 2-GOI-100-3 and documented in that procedure.

5. Add guidance for moving the Fuel Loading Chambers and a data sheet for recording pertinent information about each move.

6. Clarify the necessary requirements of the prerequisites section.

7. The recording of signal-to-noise ratio for movement of FLCs is deleted since this is now done in 2-SI-4.10.B.

8. Add requirements and recommendations for Reactor Protection System Logic.

The new procedure change did not affect any system design or functional requirement since it only directs the use of systems in accordance with their designed purpose. This procedure did not affect any of the UFSAR descriptions of testing or the specific description of the fuel loading test in Section 13.10.2. Both the UFSAR and this procedure state a purpose of safely and efficiently loading fuel to the full core size. The UFSAR description specifically states that one type of core reload is after a complete core unload where all fuel bundles have been removed from the core and subsequently reloaded with fresh fuel included and a verification performed to ensure proper configuration. This will be accomplished by the new procedure. The UFSAR and procedure acceptance criteria are also consistent, i.e., the core will be altered to reflect the design configuration while maintaining subcriticality. 2-TI-147A requires the use of FLCs which are connected to the SRM circuitry. The SRMs/FLCs are provided to monitor the core during refueling operations. Requiring one SRM/FLC in the quadrant where fuel is being moved and one SRM/FLC in an adjacent quadrant assures adequate monitoring of that quadrant during such alterations. Requiring the SRMs/FLCs to be functionally tested prior to the fuel addition assures that the SRMs/FLCs will be operable at the start of fuel loading. The response checks of the SRMs/FLCs ensures their continued operability during fuel loading.

2-TI-147A does not affect the ability of the safety systems to perform their intended function. No unreviewed safety question or Technical Specification change resulted.

1-POI-200.4 Defeating RPS and PCIS Logic for Unit 1, and 3-POI-200.4 Defeating RPS and PCIS Logic for Unit 3 Units 1 and 3

~ ~

v procedures were written to provide the instructions necessary t'hese the BFN Units 1 or 3 RPS logic and for defeating portions of the for'efeating Units 1 or 3 PCIS logic. These procedures also controlled the temporary alterations necessary to defeat this logic, and ensured that these temporary conditions (jumper placements, lifted wires, and breaker positions) are returned to normal before refueling.

These procedures affected the Units 1 and 3 RPS, PCIS, main steam system, RWCU system, RHR system, control rod drive system, primary containment vent/purge systems, secondary containment ventilation systems, and the common refueling zone ventilation, SGTS, and CREV systems.

The RPS and PCIS are integral parts of the engineered safety feature logic at BFN, and play a role in the mitigation of all of the DBAs and AOTs described in the UFSAR. However, while these procedures are in effect, the Units 1 and/or 3 reactor will be de-fueled; therefore, the only credible accident that needs to be considered for Units 1 or 3 is the fuel handling accident. The consequences of a fuel handling accident are mitigated by the existence of the secondary containment, the isolation of normal refueling floor ventilation, the initiation of SGTS', and the initiation of CREV when the event occurs as 4

k

sensed by high radiation levels. The logic from Unit 1'and Unit 3 which initiates SGTS and CREV and isolates the refueling zone normal ventilation on high radiation will not be affected by these procedures. The secondary containment structure will not affected in any way. Since the systems which mitigate the accident are not adversely affec'ted, 'the consequences of the accident cannot be increased. No unreviewed safety question or Technical Specification change was involved.

SDSP 2.1 Site Procedures Units 1, 2 and 3 t v t Numerous refinements were incorporated by this revision. Some examples are:

In section 6.4.5 implement Nuclear 'Experience Review item SQN Inspection Report 327-328/89-15 concerning annotation of Nuclear Experience Review items in procedures to prevent inadvertent deletion. This also implements STD-1.3.1, SCN-3, Revision 0.

Update to reflect numerous organizational changes. Revise list of Principal Managers in section 4.0. Change responsible organization for various procedures in Attachments E, F, and G.

In section 6.1.6 and 3.2 expand list of procedures that involve review of unusual incidents by adding reference to PMI-6.2, PMI-6.4, RCI-14, SDSP-15.10, and SDSP-27.3, and deleting superseded PMI-6.13.

Reflect paragraph number changes in STD-5.9.50 and transfer of some requirements to STD-5.9.1.

In section 4.0 delete PMP definition and add Instruction Letter, Instruction, and Standard Practice definitions by referring to section 6.2.

In section 6.1.7 expand scope of technical procedures to include monitoring consistent with SDSP-7.4.

Update and expand Attachment A to reflect current procedure system and UFSAR Figure 13.6-1. .Change Topical Report to Nuclear Quality Assurance Plan. Add Organizational Topical Report and Radiological Emergency Plan. Change T&CS-SMIs to ETU-SMIs. Add Technical Specification Interpretation Manual and Special Tests'. Delete SMGs and MMGs since they are part of Maintenance Instructions. Change Site Support to Site Programs and Support.

This procedure is consistent with sections 6.5 and 6.8 of the Technical Specifications. From a Technical Specification standpoint, its primary function is to list procedures requiring PORC review using the criteria from Technical Specification section 6.5.

0 Except for Figure 13.6-1 as noted above, this procedure was consistent with section 13.6 of the UFSAR, which lists several types of procedures. This procedure being strictly administrative in nature, no system, structure or component is directly or indirectly adversely affected. No unreviewed safety question was created and no changes to the Technical Specifications were required.

MR 898384 Temporary Power Supply During a HC Bus 2B Outage Unit 2 Under MR 898384 a temporary power supply from the 480V Turbine Building Vent Board 2A, Compartment 11B was connected to the load side of Battery Board 2 ISC Bus Breakers 825 (g24V Battery Chargers Bl-2 and B2-2) and 826 (PSS VHF Radio), via a welding receptical and a 15 KVA transformer. This temporary power supply prevented the discharge of the batteries supplied by Battery Chargers Bl-2 and B2-2 and also maintain PSS VHF radio operability during the ISC Bus 2B outage. This Safety Evaluation only addressed the installation of a temporary power supply.

The ~24V DC Power System, 480V turbine building vent board 2A, and communication system are all identified as being non-safety related in the Unit 2 Q-list.

Section 8.8.2.3 of, the UFSAR states that the g24 volt dc power system is not a safety related system and does not have a nuclear safety design basis.

Section 8.4, which describes the normal auxiliary power system, identifies no safety function for the 480V Turbine Building Vent Board System. Section 10.18.2.1.3, which describes the PSS VHF Radio System, identifies no safety function for the PSS VHF Radio System.

There are no Technical Specification requirements for any of the systems affected by this work.

For the reasons stated above, there was no impact on the margin of safety. No changes to Technical Specifications is required and no unreviewed safety question is involved.

3-GOI-100-13.18 Layup of the RHR system Unit 3 This instruction covers dry layup of Loop I and Loop II of the RHRS during the current extended shutdown period. Dry layup minimizes carbon steel corrosion to help maintain system integrity. Since the RHRS will have dehumidifiers and HEPA filters hooked up in various locations to disassembled system valves, the 0

potential exists for water. spills from these hookup points due to leaking boundary valves. During instruction implementation, identification and repair of leaking boundary valves will occur as the system is drained and components are disassembled to establish air purge paths. Periodic relative humidity sampling of air purge paths during the layup period is performed to monitor for further system inleakage.

Dehumidifier power supplies are taken from spare breakers feeding power outlets in the plant. Breakers are sized for dehumidifier operation to prevent'afety problems associated with overloaded electrical equipment.

Review of the design basis accidents in UFSAR Section 14.6 indicates no potential impact on these accidents by this new instruction since the unit will remain defueled while this instruction is in effect or until the requirements of Technical Specifications 3.5.B.9 or 10 are met for RHRS,Loop operability.

Review of Safe Shutdown Analysis calculation ND-Q2000-8815 R2 Appendix 33 for "Loss of Fuel Pool Cooling/Makeup" indicated no impact on the evaluation for the present conditions on Unit 3. Currently, there is one full core load of fuel (last day of power operation was 3/10/85) plus the spent fuel from five previous refuelings in the Unit 3 fuel pool. RHRS supplemental cooling in support of the FPC&CS has not been necessary with the amount of fuel now present in the fuel pool and the decay heat load of this fuel will continue to decrease while this instruction is in effect. UFSAR Section 10.5.4 states that the RHRS is a qualified source of makeup water for the FPC&CS and that the intertie between the RHR service water system and the RHRS can be utilized to admit raw water as makeup. Normal makeup water will still be available (from the condensate storage tank) and a standpipe and hose connection is provided on each of the two EECW headers which provides two additional qualified fuel pool water makeup sources. Therefore, no unreviewed safety question exists and no Technical Specification changes are involved.

MR 914022 Temporary Power Supply During Z&C Bus 2A Outage Unit 2 A temporary power supply was installed under this MR to supply power from a non-safety related power supply (Turbine Building Vent Board) to Battery Board 2 Panel 8 Breaker 805 (g24 Volt Power Supply System Battery Chargers Al-2 and A2-2) and Breaker 806 (Public Safety Security VHF Radio) during the 480V Shutdown Board 2A Board Outage. Z&C Bus 2A transformer was deenergized during this board outage and therefore deenergized Battery Board 2 Panel 8. This temporary power supply was required in order to prevent the discharge of the batteries charged by neutron monitoring battery chargers Al and A2 and also maintain PSS VHF radio operability during Z&C Bus 2A outage. This safety evaluation only addressed the installation of a temporary power supply.

The 24V DC Power System, 480V Turbine Building Vent Board 2A, and Communications system are all identified as being non-safety related in the Unit 2 Q-list.

Section 8.8.2.3 of the UFSAR states that the 24V DC power system is not a safety related system and does not have a nuclear safety design basis.

Section 8.4, which describes the normal auxiliary power system, identifies no safety function for the 480V Turbine Building Vent Board system. Section 10.18.2.1.3, which describes the PSS VHF Radio System, identifies no safety function for the PSS VHF Radio System.

There are no Technical Specification requirements for any of the systems affected by this work.

For the reasons stated above, there was no impact on the margin of safety, and no unreviewed safety question involved. There were no Technical Specification changes.

MR 914022, Rl Temporary Power Supply to Battery Board 2 Panel 8 Unit 2 A temporary power supply was installed under this MR to supply power from RMOV board 2A compartment 8A to Battery Board 2 Panel 8 during the 480V Shutdown Board 2A outage. This temporary power supply was required in order to prevent the discharge of the batteries powered by neutron monitoring battery chargers Al and A2, to maintain PSS VHF radio operability and to maintain power to the fire protection panels fed from Battery Board 2 Panel 8. The power for this temporary connection was supplied from 480V RMOV Board 2A compartment 8A.

This compartment is a spare compartment on the board. Because of the present plant configuration (all units defueled) and the non-,.required breakers being open on the I&C bus, the additional load had no adverse affect on the 480V RMOV board. The 100 amp spare breaker adequately protected the RMOV board from electrical faults. Therefore, no electrical loading problems will be encountered. No safety related equipment was supplied by this temporary power supply. These systems still functioned as required to mitigate the events

-applicable to the plants present configuration. As a result, no unreviewed safety question or Technical Specification change was involved.

WO 09-02037-00 Temporary Power Supply to I&C Bus 3A Unit 3 A temporary power supply was installed under this WO to supply power from RMOV

.'oard 3A compartment 9A to Battery Board 3 Panel 8 (Z&C Bus 3A) during the 4KV Shutdown Board 3EA and 3EB outage. This temporary power supply was required in order to prevent the discharge of the batteries charged by neutron monitoring battery chargers Al-3 and A2-3, to maintain RRRMS functional and to

maintain power to the fire protection panels fed from Battery Board 3 Panel 8 (I&C Bus 3A). The temporary source, 480V RMOV Board 3A, was protected from any faults by the 100 amp spare breaker. No safety related equipment was supplied by this temporary power supply. These systems functioned as required to mitigate the events applicable to the plants present configuration. No unreviewed safety question or Technical Specification change was involved.

1-GOI-100-13.18 Layup of the RHR System Unit 1 1-GOI-100-13.18, "Layup of the RHRS," implements dry layup of Loop I of the RHRS during an extended shutdown period. Dry layup minimizes carbon steel corrosion to help maintain system integrity. Layup is established by isolating (placing Hold'Orders on all boundary valves) and then draining Loop I. To set up air purge paths, check valve internals will be removed or wired open, dehumidifier(s) connected to the Loop I RHR Drywell Spray Outboard valve (FCV-74-60), and HEPA filters connected to the RHR Pump A&C suction drain valves, RHR Drain Pump A suction check valve, Flush & Fill check valves, and three test connections. Valve components removed from the system will be handled and stored in accordance with SDSP-16.19 (Maintenance/Modifications Material Control). Drain valves are isolated prior to turning on the dehumidifier(s). Dehumidifier(s) are of the solid dessicant type having a rotary dessicant bed. System status control is maintained for valve, panel, and electrical lineup checklists using PMI-12.15. Reassembly of all valves and then releasing the system from layup status control is provided for in the "Restoration" section of this instruction.

The decision to use Loop II of unit 1 RHRS to support power operation of unit 2, the removal of fuel from the unit 1 reactor vessel, and the small decay heat load of the fuel presently in the fuel pool, alleviates the need to maintain Loop I of unit 1 RHRS operable. Review of the design basis accidents in UFSAR section 14.6 indicates no impact on these accidents by this new instruction since the unit will remain defueled while this instruction is in effect.

Review of Safe Shutdown Analysis calculation ND-Q2000-8815 R2 Appendix 33 for "Loss of Fuel Pool Cooling/Makeup" indicates no impact on the evaluation for the present conditions on unit 1. Currently there is one full core load of fuel (last day of power operation was March 19, 1985) plus the spent fuel from five previous refuelings in the unit 1 fuel'pool. RHRS supplemental cooling in support of the FPC&CS has not been necessary with the amount of fuel now present in the fuel pool and the decay heat load of this fuel will continue to decrease while this instruction is in effect. No Technical Specification change or unreviewed safety question resulted.

SAFETY EVALUATIONS FOR PROCEDURE REVISIONS

PMI-15-10 Tracking of Limiting Conditions for Operations Units 1, .2 and 3 t t The change was to a procedure which is not mentioned in the UFSAR. The-procedure directed and tracked actions that are necessary to ensure compliance with Technical Specifications when Technical Specification equipment becomes inoperable. The procedure revision did not involve changes, tests, or experiments to any structures, systems or components described in the UFSAR or procedures mentioned or outlined in the UFSAR. However, the procedure did involve a detailed administrative program to assure Technical Specification

. equipment LCOs, compensatory actions and work items affecting system and/or component operability are adequately tracked and resolved. Tracking items affecting an operability declaration for Technical Specification equipment assures sy'stematic resolution of those items to the degree that there is no open issue which would place the plant in violation of a LCO. As such, the administrative program imposed controls which provide documentation and assurance that plant operations can be continued in an atmosphere of enhanced safety. No design basis accidents which are evaluated in the UFSAR were found to be affected by this revision. Neither were any UFSAR identified failure modes found to be applicable to this procedure revision. This procedure revision is administrative in nature and provided controls which enhance compliance with Technical Specification requirements. No unreviewed safety question or Technical Specification change resulted.

EOI-3 Secondary Containment and Radioactive Release Control Unit 2 2-EOI-3 addressed a broad spectrum of events, including those within the design basis events and beyond the design basis events. 2-SOI-3 is based on the PSTG. The PSTG is, based on the BWROG EPG, Revisions 3 and 4. Both the PSTGs and the EPGs have received NRC review and approval. This procedure did not create the possibility for an accident of a different type than previously analyzed in the UFSAR. This procedure is used only in response to an accident and these changes make the procedure more user friendly without significantly changing the technical basis.

2-EOI-3 specifies the correct technical guidance in the proper format to accomplish Secondary Containment'nd radioactive release control. This procedure was verified and validated in accordance with plant procedures

'NI-12.8 and PMI-12.9. Therefore, the proce'dure does not create the possibility of a different malfunction since the equipment is to be operated in the same manner. No unreviewed safety question or Technical Specification change resulted.

4 EOI-2 Primary Containment Control Units 1, 2 and 3 t

result of plant modifications, validation t'hese changes were made as a comments from Operations personnel to improve procedural execution and as a result of an NRC review of the EOIs. The procedure provides written directions to accomplish the task of primary containment control and these changes were made to enhance the procedure. Thus the probability that equipment will be improperly operated is decreased and the probability that an equipment malfunction will be promptly identified is increased. The procedure was verified to assure that it provides the correct technical guidance and that it is properly human factored. The procedure was validated to assure it it that it is usable, that it will accomplish purpose, and that is compatible with the control room and plant equipment. No unreviewed safety question or Technical Specification change resulted.

O-SI-4.2.D.4 Radiological Liquid Effluent Radwaste Monitoring Instrumentation Units 1, 2 and 3 t v This SI provides for the calibration and functional testing of the Radiological Liquid Effluent Radwaste Monitoring Instrumentation (O-FIT-77-90, and 0-FR-77-60). This instruction partially satisfies the requirements specified in Technical Specification Tables 3.2.D and 4.2.D, and Liquid Radwaste Effluent Flow Rate (77-60 loop).

This instruction change consisted of a general revision to include the calibration of a different model FIT and FR.

The functional portion of this instruction consists of calculating tank level decrease and comparing the calculated release to the totalizer of the FIT.

Calibration of radwaste effluent discharge flow loop components will not cause faults or adversely affect any safety related components. The components are consistent with the design of the radwaste system and potential failures associated with the components are no different than those already considered in the UFSAR. The proposed calibration did not involve an initiator or failure not considered in the UFSAR whose effects were not bounded by other events analyzed in the UFSAR, nor did it increase the probability of malfunction of equipment or involve a newly discovered malfunction of equipment, previously thought incredible, to the point where it becomes credible. No unreviewed safety question or Technical Specification change resulted.

EOI Program Manual Rl Units 1, 2 and 3 The EOI Program Manual provides the technical basis for the symptom-based EOIs. It consists of PSTG; Appendices A, B, C, and D to the PSTG; and the Deviations Cross Reference.

The PSTG is developed from the generic EPGs and provides BFN specific action steps, cautions, entry conditions, and action limits for controlling reactor parameters such as level, pressure, and power; containment parameters such as drywell temperature, primary containment pressure, and suppression pool level and temperature; secondary containment levels, temperatures, and radiation levels; and radiation release rates in order to mitigate an entire spectrum of events including less than design basis, design basis, and beyond design basis. These changes to the EOI Program Manual did not impact the Technical Specifications. They provided technical guidance for operations of equipment newly installed, and new action levels based on the modification that changed the post accident flooding range level instrumentation scales and reference zero. By providing the correct technical guidance, the consequences of an accident are mitigated. These changes primarily addressed events beyond the design bases and so are based on the EPGs revision 3 or revision 4. These changes to the EOI Program Manual did not provide any directions for equipment operation that compromises any margin of safety in the Technical Specifications. No unreviewed safety question or Technical Specification change resulted.

EOI Program Manual R2 Units 1, 2 and 3 t t t The EOI Program Manual consists of PORC reviewed, controlled documents containing the technical source material used in the process of developing EOIs. These documents include the Appendices A, B, C, and D to the PSTG, and the Deviations Cross Reference. The PSTG is developed from the Boiling Water Reactor Owners Group EPG Revision 3. Revision 4 of the EPG also was used in developing the PSTG. Both revisions have been reviewed by NRC and approved for implementation. Although portions of 'Revision 4 have been used, Revision 3 is the primary source for the PSTG and the deviations from the generic guidance are compared to Revision 3. The PSTG is developed from the generic EPG and provides the specific actions, cautions, entry conditions, and action limits for controlling reactor parameters such as Drywell temperature, Primary Containment pressure, Suppression Pool level and temperature; Secondary Containment levels, temperatures, and radiation levels', and radiation release rates in order to mitigate an entire spectrum of events, including less than Design Basis, Design Basis, and beyond Design Basis. All of these documents constitute the EOI Program Manual with the purpose to provide the technical input necessary to develop and document the basis of the symptom based EOIs.

Changes to the EOI Program Manual provided the technical information necessary to prepare symptom based EOIs. These changes did not affect the Radwaste system, nor constitute a special test or experiment. These changes did not change any system design or functional requirements, nor change any text, tables, graphs, or figures presented in the UFSAR except for the ADS.

The changes to the EOI Program Manual resulted in changes to the EOIs that differ from the description of operation of the ADS. Change RC/L-2.2 was to inhibit the ADS rather than resetting the timer. This was done because a modification has been performed on the ADS which installed the inhibit switches. The basis for this step contained in the EPG Revision 4 permits this change to be made. The ADS is inhibited rather than resetting the timer if the operator determines that reactor water level cannot be maintained above for this is that ADS actuation the ADS initiation setpoint. The reason imposes a severe thermal transient on the reactor vessel and may complicate efforts to restore and maintain reactor water level as specified in step RC/L-2. In certain cases (e.g., HPCI/RCIC available but LPCI/LPCS injection valves closed and control power not available) ADS actuation may directly lead to loss of adequate core cooling and subsequent core damage. Further, the conditions assumed in the design of the ADS are not present (e.g., no operator action for ten minutes after event initiation) when the actions are being carried out. Finally, an operator can draw upon much more information than is available to ADS logic and can better judge when to depressurize the reactor.

It is not appropriate to tie an operator up by requiring him to reset the timer at this point. Subsequent steps in the PSTG tell him when to depressurize. Emergency reactor depressurization (manual ADS) is called for when the reactor water level drops to top of active fuel in Cl. Per UFSAR Figure 6.5-17 for core response to ADS with 0.05 feet squared liquid line break, the level is seen to be slightly below top of active fuel when ADS initiates. Therefore, the manual action is taken in time to not invalidate the analysis conclusions for intermediate line breaks. This is in accordance with the EPG Revision 004 guidance. No unreviewed safety question was created and no Technical Specification change resulted.

SAFETY EVALUATIONS FOR SPECIAL OPERATING CONDITIONS

SE 00-8912-003 Revision 2 . Unit 2 Core Unload The nuclear fuel that was located in the BFN unit 2 reactor was completely unloaded as specified in Fuel Trarisfer Operation 8BFN-2-35. At some point, one or more Fuel Loading Chambers (FLC) will be used in lieu of Source Range Monitors (SRM) A and B in order to maintain an operable detector in each core quadrant where fuel is being moved and in an adjacent quadrant. The FLCs are utilized since removal of additional fuel assemblies will uncover the SRMs.

The remaining fuel was removed in an order intended to maintain adequate and redundant neutron monitoring.

The defueling activities were conducted in accordance with BPK unit 2 Technical Specifications 3.10.B/4.10.B. All control 'rods were inserted and electrically disarmed for the duration of the operation in accordance with TS 3.10.B.2. This eliminates the creditability of the rod drop DBA. To ensure adequate and redundant neutron monitoring to detect changes in the reactivity condition of the fuel, BFN also administratively required the implementation of the more stringent requirements on neutron monitoring that apply to fuel loading (i.e., Technical Specification 3.10.B.1/4.10.B.l, which requires two SRMs (FLCs) to be operable, one in and one adjacent to any quadrant where fuel is being moved) Under these conditions, all fuel moves during core unloading

~

will reduce reactivity. The analyzed fuel bundle drop DBA is a possibility in the conduct of fuel transfer operations with the exception of the reactivity level being present in the fuel. The DBA assumes worse case of 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months decay time', in this defueling, there is four years and nine months decay time.

Therefore, considerably less inventory of fissionable products exists than analyzed for in the UFSAR fuel drop accident DBA. This core unload did not require a Technical Specification change or involve an unreviewed safety question.

WP 2453-90, Revision 1 Installation of Temporary Jumpers to 2-LIS-77-1A and 2-LIS-77-1B Unit 2 v t'P 2453-90 Rl installed a temporary three-conductor jumper to maintain operation of 2-LIS-77-1A and 2-LIS-77-1B while work was performed to reroute conduit associated with these level switches and other equipment as depicted in DCN W9962. This jumper, allowed Operations personnel to monitor the water level in the unit 2 drywell floor drain sump and manually operate temporary sump pumps installed in this sump to control water level. This alteration consisted of a three conductor jumper cable in cables 2PL2337 and 2PL2359 between JB3004 and JB3010. Interim measures to provide level indication during jumper installation included a physical inspection of the sump level every four hours. This alteration will be closed prior to fuel load per WP 2453-90. This activity did not affect any design basis parameters (i.e.,

temperatures, pressures, flow, heat removal, etc.) listed in UFSAR Chapter 9.2 and 10.16 for the Radwaste Systems. This activity did not alter any information, nor change initial condition, assumption, etc., in UFSAR Sections 1'

14.5 (Analysis of Abnormal. Operational Transients), or 14.6 (Analysis .of Design Basis Accidents). The credible failure modes for this alteration were identical to the modes under normal configuration. The jumper cable met the same requirements as the existing cable. Operations and/or Nodifications personnel continued to monitor the level in the drywell floor drain sump and act accordingly. No unreviewed safety question or Technical Specification change resulted.

2-SOI-57C-01 Temporary Power Supply to Battery Board 2 Unit 2 A temporary power supply was installed under this SOI to supply power from RMOV Board 2A compartment 8A to Battery Board 2 Panel 8 during the 480V Shutdown Board 2A outage. This temporary power supply was required in order to prevent the discharge of the batteries supplied by neutron monitoring battery chargers Al-2 and A2-2, to maintain PSS VHF radio operability, to maintain power to the fire protection panels fed from Battery Board 2 Panel 8 and to provide power to 2-PX-64-50 which supplies power to 2-LT-64-54 (torus water level), 2-PT-64-50 (drywell pressure) and 2-PT-64-51 (torus pressure).

The installation of the temporary power supply required a temporary outage on the affected systems during -its installation and removal. Appropriate operating procedure action was taken to insure these temporary outages were acceptable. The only credible failure mode associated with this change was the loss of power to B'attery Board 2 Panel 8, which would remove power from Battery Changers Al-2 and A2-2, the PSS VHF radio system, the fire protection panels fed from the panel and., an instrument power supply (PX-64-50) in the Backup Control Panel. The temporary power supply is diesel generator backed just as the normal supply.

Neither the battery chargers nor the PSS VHF radio system has any effect on the mitigation or initiation of a design basis accident or event. Therefore, losing them would not affect nuclear safety.

With Unit 2 defueled, the requirement for Power Supply PX-64-50 which powers instrument loops 2-LT-64-54 (torus water level), 2-PT-64-50 (drywell pressure) and 2-PT-64-51 (torus pressure) did not apply. However, these instrument loops would still function properly since the equipment would remain energized from a diesel generator backed power source. Since the plant had all three units defueled, the requirement to shut down the plant during a fire did not apply. However, the fire protection system would still function properly since the equipment would remain energized from a diesel generator backed power source. This SOI did not remove any required safety related equipment from service and operating procedure actions were followed whenever required-equipment was out of service. The fire protection system would still perform its required design function for the plant's present configuration. No unreviewed safety question or Technical Specification change resulted.

SA B%HSR 90053 Plant Configurations During Multiple Modification Activities Units 1, 2 and 3 t v The design basis accident evaluated in the Safety Analysis Report which may be affected by modification activities causing interim changes to the plant configuration is a LOCA. This design basis accident is evaluated for the current plant conditions:

a. all reactors defueled,

b. all fuel assemblies stored in the spent fuel pools,

c. no fuel handling activities or operations over the fuel pools permitted,

d. the plant shutdown for over five years,

e. instrumentation specified in Technical Specification 3.2/4.2 operable,

f. spent fuel water temperature and chemistry within Technical Specification 3.10/4.10 limits, and

g. fire protection systems operable in accordance with Technical Specification 3.11/4.11 This safety evaluation is invalid for modification activities if current plant conditions a through c change. All modification activities referencing this safety evaluation will cease, through work plan instructions, if plant conditions a through c change. If conditions e through g change, the SOS, through plant work control procedures, will evaluate modification activities on a case by case basis to determine if the modification activity can continue.

Permanent modifications made to the plant configuration require separate safety evaluations and PORC review prior to modification implementation as per current plant procedures.

The credible failure mode which may occur during modification activities causing interim changes to the plant configuration is fuel pool draining.

There are no credible accidents of a different type than those evaluated previously in the UFSAR. Restricting activities which have the potential to drain the spent fuel pools precludes the risk of a LOCA. When secondary containment is inoperable,.Technical Specification 3.7.C.2 does not permit fuel handling activities or operations over spent fuel pools which precludes the risk of a refueling accident.

New possible malfunctions to equipment important to safety are not introduced by interim changes to the plant configuration under the current plant conditions. Restricting activities which have the potential for draining the spent fuel pool and operations over the fuel pool precludes the risk of LOCAs or Refueling Accidents. No new induced equipment malfunction pathways are introduced by interim 'changes to the plant configuration which have not been previously evaluated in the UFSAR. No unreviewed safety question or Technical Specification change resulted.

Technical Specification Interpretation 90-01 EECW Operability Units 1, 2 and 3 Technical Specification Interpretation 90-01 specifies the requirements for operable EECW headers during shutdown conditions with all three units defueled. This Technical Specification Interpretation allowed the removal of one of the two EECW headers indefinitely to accommodate modifications and maintenance. This required isolation of some cooling loads from the EECW system to ensure adequate cooling is provided for the required loads.

The only design basis accident possible with all three units defueled is the loss of offsite power and fuel handling accident as described in UFSAR Section 14.6.4. None of the assumptions or results from this accident are altered by requiring only one of the two EECW headers to be operable. All the systems required to operate in response to this accident (SGTS, Reactor Building Isolation, Secondary Containment, Control Bay HVAC, Reactor Building Radiation Nonitoring) are not adversely affected by this activity and will perform as analyzed in the UFSAR.

Technical Specification basis, 3.5.C states that all components requiring EECW can be adequately supplied cooling water if either header is operable. It also states that two RHRSW pumps can supply full flow to all essential EECW loads for any abnormal or post-accident situation. This activity maintained one header operable and limited the loads on the system such that one pump can supply full flow for essential loads required to mitigate a fuel handling accident with all units defueled. Therefore, the margin of safety as defined in Technical Specification bases has not been reduced. No unreviewed safety

'question was involved and no changes to the Technical Specifications were required.

SE 67-9002-006 RO Backwashing EECW Strainers Units 1, 2 and 3 ft v The proposed activity allowed an EECW strainer to be maintained in an operable condition after the failure of the automatic backwash feature via the implementation of the requirement to manually backwash the strainer once every 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The basis for Technical Specification 3.5.C states that all components requiring EECW can be adequately supplied cooling water if either header is operable. It also states that two RHRSW pumps can supply full flow to all essential EECW loads for any abnormal or post-accident situation. The Technical Specifications also specify the minimum EECW pump requirements for continued operation. The special requirements assured that the strainer remains operable even without its continuous backwash, which ensured that the pump aligned to the strainer remains operable. Therefore, the required margin of safety was maintained. The proposed activity contained a special requirement needed to maintain an EECW strainer and pump operable without the continuous backwashing of the strainer. This special requirement to manually backwash the strainer every 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months was implemented to compensate for the loss of the continuous backwash. These measures were implemented by incorporating them into a change to,GOI-300-1 that required an AUO to manually open the strainer backwash drain valve and rotate the strainer every shift if the automatic backwash is inoperable. As stated in calculation ND-f2999-880160, a 48, hour manual backwash would be adequate to maintain the strainers in an operable condition. Therefore, these measures ensured that the EECW strainer did not clog and that EECW performed its safety function.

No unreviewed safety question or Technical Specification change was involved.

SAFETY EVALUATIONS FOR SPECIAL TESTS STs 88-16 and 88-17 Load Acceptance Test for DGs A and B Unit 2 t

t'his ST delineated the necessary system alignments and operating sequences required to obtain test data which was used to perform a stability analysis on the diesel generator system response during a worst case accident loading sequence. The test was performed with the reactor in a cold shutdown condition without fuel in the reactor vessel. Thus, any adverse effects on plant safety due to having fuel in the core were eliminated.

Electrical loading of the DG was achieved using normal plant loads on the Shutdown Boards. Systems powered from the Shutdown Boards during this test were aligned and operated in accordance with approved plant operating instructions. Pre-selected loads were specified by NE in order to achieve the required loading for the stability study. Maximum loading of the DG was within the design limits of the DG for the entire duration of the test. Since the DG was loaded only by systems normally powered from the DG, there were no adverse effects on safety to the DG while performing this test.

The RHR discharge headers for Loop I and Loop II were cross-connected during performance of this .test by opening Crosstie Valve FCV-74-46. This abnormal RHR lineup simulated worst case loading of RHR Pump 2A by allowing it to discharge at its maximum design flow rate of 12,000 gpm through both test return lines on the two RHR discharge headers. (Normal flow through one test return line is 9,000 gpm.) Flow through the RHR pump is restricted to 12,000 gpm design flow by an orifice installed on the discharge of the pump. These instruction procedures provided additional RHR pump protection by requiring the operator to monitor and throttle RHR pump flow at or below 12,000 gpm during performance of this test. For the second portion of the test the RHR discharge header crosstie valve was closed and RHR pump flow was reduced to 9,000 gpm maximum; thus, the RHR system was operated in its normal mode during the second portion of testing. Since the RHR pump was operated within its design limits, no adverse safety effects were incurred on this component.

During performance of this test, all equipment was operated within design limits and in a manner in which the equipment was intended to operate. There were no system modifications or lineups specified by the test that would cause equipment to operate outside any design limits. No unreviewed safety question was created and no Technical Specification change resulted.

ST 88-35 Testing Pressures in Stack Dilution Ductwork During SGTS Operation Units 1, 2 and 3 t' t ST 88-35 was performed to determine what positive pressure exists in the Stack Dilution Air Fan ductwork when the SGTS was in operation and the Stack Dilution Air Fans and Filter Cubicle Exhaust Fans were shutdown. This information was used in the resolution of CARR BFP 880304 Rl which documented a positive pressure in this ductwork. In this test two trains of SGTS were

'I 0

started with a minimum of .the Unit 2 and 3 Stack Dilution Air Fans in ~

operation. The flow produced in this state was recorded. At this point all units Stack Dilution Air Fans and Filter Cubicle Exhaust Fans were shutdown.

The flow was then recorded again. The third train of SGTS was started and the positive pressure produced in the exhaust duct of the Stack Dilution Air Fans was measured. The systems were then returned to a normal operating state.

The Offgas Dilution fans and Filter Cubicle Exhaust fans are only required when discharging radioactive materials through the Offgas system or when a unit is in operation in order to dilute the amount of hydrogen present in the offgas exhaust to the stack and to exhaust the filter cubicle area. They are not required to mitigate an accident. Shutting down these fans in this 'ST did

'not affect safety since the special requirements did not allow this test to be performed with any unit in operation or during radioactive releases. The only equipment affected by this test was the SGTS, Stack Dilution Air Fans and Filter Cubicle Exhaust Fans. This ST only manually started and stopped this equipment in the control room. This operation created no new methods of equipment failure which should have been considered in previous UFSAR analysis. No unreviewed safety question was created and no Technical Specification change resulted.

ST 89-04 Measuring RHRSW Flows and Pressures During RHR Heat Exchanger Operation Unit 2 t v This ST measured flows and pressures in the RHRSW system by lining up the RHRSW pumps'o the unit 2 RHR heat exchangers. The purpose of this test was to provide a measurement of system performance to Nuclear Engineering. This information was requested as part of the detailed design analysis of system requirements that was performed to resolve CAQR BFP 880483 (written to address low RHRSW flow through the RHR heat exchangers during the performance of surveillance instruction 2-SI-4.5.C.1(3)). The equipment and systems affected by this special test were not operated outside of their normal expected design parameters. All system and equipment alignments were as described in existing plant instructions and as described by the UFSAR. RHRSW system and equipment operability as required by Technical Specification 3.5.C was not affected by this ST. No unreviewed safety question was created and no Technical Specification change resulted.

ST 89-06 Evaluation of Pall Rigimesh Filter Elements in the Radwaste Floor Drain and Waste Collector System Units 1, '2 and 3 t t The purpose of this test was to evaluate the operation of the Pall Rigimesh filter elements in the radwaste floor drain and waste collector system. Thes'e elements replaced the existing Croll-Reynolds elements presently installed in the waste collector and floor drain filter vessels. Use of the precoat tank by-pass line (to reduce precoat rate) and body-feed system should reduce resin 0

generation and increase processing efficiency. The filter aid and cation floe feed system injected various filter aid and cationic polymer products into the filter inlet piping. Testing of the Pall Rigimesh filter elements in the radwaste system in no way affected the safety of the radwaste system. The function and flow path of the radwaste system'emained unchanged. The design criteria for the Pall Rigimesh element was greater than or equal to the design criteria of the Croll-Reynolds elements. Although Figure 9.2-2 of the UFSAR states that the existing elements'iltering area is 200 feet2 and the maximum differential pressure is 50 psid, the new elements had a filtering area of 192 feet and a maximum differential pressure rating of 100 psid.

Operating within the design limits of the Pall Rigimesh elements during ST 89-6, though above the values identified on UFSAR Figure 9.2-2 which are based on the Croll-Reynolds elements, did not invalidate any safety analysis of the radwaste system. Therefore, the design criteria of the affected components as well as the filter elements was maintained. As stated in UFSAR Section 9.2.6, a loss of tank contents within the Radwaste Building would be contained within the Radwaste Building designed to withstand a Design Basis Earthquake.

Testing of these elements inside the Croll-Reynolds vessel did in no way affect the seismic qualifications of the Radwaste Building or a loss of tank contents. The Pall Rigimesh elements were designed to improve the processing efficiency of the radwaste system. Therefore,, the probability of tank overflow due to inefficient processing was reduced. No 'unreviewed safety question was created and no Technical Specification change resulted.

ST-89-07 Main Offgas Stac'k Backflow Test Units 1, 2 and 3 ST-89-07 was performed to determine the amount of backflow through the Stack Dilution Air Fan and Filter Cubicle Exhaust Fan ductwork when the SGTS is in operation, and the'Stack Dilution Air Fans and Filter Cubicle Exhaust Fans are shutdown. It also determined the amount and possible cause of backflow into SGTS with the Stack Dilution Air Fans and Filter Cubicle Exhaust Fans in operation and SGTS shutdown. SGTS is required to mitigate the consequences of various accidents by filtering the contaminated atmospheres from the accidents. The special requirements limited this test to the current Unit 2 Cycle 5 outage. While in this outage, the only credible accident was the fuel handling accident. The special requirements also limited this test to when there were no activities over spent fuel pools and opened reactor wells containing fuel. These requirements essentially eliminated the need for SGTS. The only equipment operated in this ST that is described in the Technical Specification was the SGTS. The Stack Dilution Air Fans, Filter Cubicle Exhaust Fans and Steam Packing Exhausters are,not described in the Technical Specification bases. The Mechanical Vacuum Pumps were not manipulated in this"test, they were only verified to not be in operation. The SGTS was placed only in modes of operation that are specified in the UFSAR and Technical Specifications which assured that all requirements specified were met. No unreviewed safety question was created and no Technical Specification change was required.

ST-89-10 Evaluation of Norton Ceramic Filter in the Liquid Radwaste System Units 1, 2 and 3 The purpose of this test was to evaluate the operation and efficiency of the ceramic filter in either the floor drain sample tank or the waste sample tank. This filter proved beneficial in removal of organic and ionic impurit'ies in the Radwaste System. Liquid waste was recycled through the affected sample tank and filter skid. Concentrated liquid waste (bleed) from the filter was routed to the Radwaste Condensate and Waste Phase Separators via pressure rated rubber hose at a flow rate of approximately 25 gpm. Clean permeate was returned to the affected sample tank at a flow rate of approximately 25 gpm. The -total effluent (permeate and/or concentrate) flow was approximately 25 gpm. After draining the affected tank, the skid inlet was connected to the blank flange on the tank bottom and the clean permeate was connected to the blank flange on the top of the tank via pressure rated rubber hoses. Normally, the only methods to recycle the affected sample tanks are through existing recycle pumps and piping. This test defined another method to recycle these tanks. There was no impact on the margin of safety for this ST. The filter was designed, installed, and operated to meet all Radwaste Safety Requirements. The filter increased the particle removal rate resulting in a decrease to effluent radioactivity levels. RADCON control of all operations were maintained. As a result, no Technical Specifications changes were required with no unreviewed safety questions involved.

ST 90-01 Hi-Pot Testing Units 1, 2 and 3 Due to the discovery of damage to cables as a result of cable pullbys during installation at Watts Bar Nuclear Plant (CARR WBP890331), there was concern for the existence of similar damage at BFN. This test procedure provided instructions for performing in-situ High Potential Tests on low voltage cables to identify if the postulated damage exists. All equipment was disconnected from the cables during the testing operation. Also, all equipment and cables were tested, under existing procedures used for post maintenance testing, before being returned to service. Hi-pot testing is a standard practice performed throughout the industry on installed medium voltage cables, and by cable manufacturers on both medium voltage and low voltage cables shortly after their manufacturing process is complete. Since the voltages applied during the test were based on industry guidelines, this test did not have adverse effects on the cables which pass the test. The acceptance criteria for passing the test was appropriate for identifying the suspect, damaged cables. The applied test voltages were high enough to locate the suspected damage and low enough to prevent damage to cables in satisfactory condition.

No unreviewed safety question or Technical Specification change was involved.

ST 90-11 Rod Worth Minimizer Unit 2

~ ~

t v ST 90-11 was implemented to provide a baseline test for all functions of the RWM system. The RWM system has been in use for a number of years, but has undergone software and hardware modifications without sufficient documentation to allow the present system configuration to be determined exactly. The purpose'f this ST was to test actual system operation against the description found in the RWM users manual (GEK-39439D), and to identify and document any discrepancies between the two. The method employed by this ST was to withdraw control rods as would be done for a normal startup and test the various functions of the RWM along the way. Since the unit 2 reactor vessel has been defueled, the reactor mode switch may be positioned as desired and rods withdrawn without any need for neutron monitoring, containment, or other equipment operability requirements which are normally needed for a plant startup. This ST also allowed for open vessel scram time testing to be performed after the RWM testing was completed, and was used to provide a test of the ARI system to re-insert the control rods at the end of the ST. This ST allowed the bypassing of certain reactivity control systems so that the RWM may be tested without interference. However, for the operating conditions under which this test was conducted (head removed, vessel defueled), there were no possible accidents which would require reactivity control interlocks since no fissile material was involved. No other systems were affected by the implementation of this ST, so plant design and operation was as described in the UFSAR. Control rod withdrawal for the ST proceeded as is normally done for a reactor startup to verify the correct functioning of the RWM. The safety margins which form the bases of Technical Specifications on reactivity controls are to protect against fuel cladding barrier overstress resulting from high heat generation rates in an operational transient (such as for the RMCS rod blocks and RPS automatic scrams) or against high peak fuel pellet enthalpy during a rod drop accident (RSCS and RWM rod blocks). Because the vessel was defueled, these safety margins were not a concern since they all deal with nuclear fuel in an operating environment. There were no other safety margins defined in the Technical Specification bases which were affected by this ST. No unreviewed safety question or Technical Specification change resulted.

0' SAFETY EVALUATIONS FOR TENPORARY ALTERATIONS TACF 2-90-001-79, RO Temporary Installation of a Support Bracket to the Refuel Bridge Crane Unit 2 This TACF allowed installation of a bracket plate and clamps with associated aluminum shim plates to the vertical posi.tion indicator of the Unit 2 refuel bridge crane. The bracket plate was the mounting surface for'he clamps that stabilized the flexible shaft for the vertical (z-axis) position indicator.

This TACF-was required to ensure the repeatability of readings by the vertical position indicator during refueling operations.

1 Section 10.4.6 of the UFSAR currently states that platform operations are controlled from a walkway and from an operator station on the trolley. The platform contains a position-indicating system that indicates the position of the fuel grapple over the core. The original position indicating system was a three axis indicator system supplied by the crane manufacturer. Currently, a manual line up and sighting system for Z and Y axis position indication is utilized by the operator. This involves position marks on the crane that are aligned with posi.tioning marks on a yardstick type device temporarily secured to the handrails adjacent to the fuel pool. Therefore, the position-indicating system is the composite of the sighting system and the vertical position shaft.

This TACF was a temporary modifi.cation and was not a new procedure or instruction, or a change to any procedure or instruction. Neither the probability of the occurrence or the'onsequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the UFSAR was increased or created.

Therefore, no unrevi.ewed safety question was involved and no changes to the Technical Specifications were required.

TACF 2-90-002-303 Personnel Access Restricting Cage Over Unit 2 Drywell Equipment Hatch Unit 2 t'

t'his TACF involved the installation of a wi.re cage around the drywell equipment Northwest hatch on unit 2 at Elevation 565.0 at Column Lines 10 and P. This temporary change increased ventilation inside the drywell and prevented entry into the drywell except through controlled points. During the current outage period, posi.tive security must be provided. This will minimize manpower by not requiring a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months security watch. The Unit 2 reactor was in an extended outage with structural modifications being required in the drywell. In this condition, primary containment was not required and the change did not adversely affect the integrity and isolation function of the primary containment. The cage has been verified by analysis and will not fall on any existing safety related installation in order to comply with the Class I criteria currently required by the UFSAR. These changes di.d not alter the design, function or method of performing the function of any system, structure or components described in the UFSAR.

The changes for the TACF did not require procedure/instruction changes and did not directly or indirectly affect the plant radwaste system. This TACF did not have any impact on the Plant Safety Analysis and therefore, no unreviewed safety question resulted. No change to the Technical Specifications was required.

TACF 2-90-003-24 Temporary Cooling Water to Drywell Cleanroom ACU Unit 2 The proposed change involved providing cooling water to the temporary drywell clean room ACU. The cooling water was provided by the Raw Cooling Water system. The only events associated with. the change were loss of Fuel Pool cooling in the event that Raw Cooling Water was lost. Additionally, EECW is the Class I backup to RCW and is initiated on low RCW header pressure. The Raw Cooling Water System was not adversely affected by the proposed change and the RCW contribution 'to Fuel Pool Cooling was therefore not degraded. The only type of accident which might be created by this TACF would be related to fuel failures resulting from loss of Fuel Pool Cooling. However, loss of RCW has previously been evaluated in the UFSAR; therefore, the possibility of a new type of accident has not been created. No unreviewed safety question or Technical Specification change resulted.

0640k 0

0

SAFETY EVALUATIONS FOR PLANT MODIFICATIONS DCN B0044B and 45B Changeout of Units 1, 2 and 3 Diesel Generators Field CB With Molded Case Switches Units 1, 2 and 3 DCN B0044B and B0045B replaced the 100 AMP, Type TED DG field circuit breakers in each of the Unit 1, 2, and 3 DG field circuits with Class lE, 100 Amp, Type TED molded case switches with shunt trip and increased the wire size for the DG field exciter circuits from number 4 AWG to number 2 AWG. These changes were made-because one of the field circuit breakers tripped during DG load acceptance testing when the generator was loaded to its 7-day rating of 2950 KW at 0.8 PF. Subsequent testing revealed that the temperature inside the cabinet rose to greater than 50'C and the elevated temperature contributed to the breaker tripping at a current lower than its rated current. The molded case switch was installed in the same electrical control cabinet which contained the field circuit breaker. However, the location inside the cabinet was changed due to installation interferences. The replaced DG field exciter cables were rerouted inside the cabinet to accommodate the new location of the molded case switch. No change was made to the generator field windings. The effect on the DG system was limited to the field exciter circuits. A molded case switch has no thermal trip elements to cause a thermal trip. Replacing the field exciter cables with larger cables and replacing the circuit breakers with molded case switches had no affect on the function, operation or qualification of any component or system required to ensure nuclear safety.

No unreviewed safety question was created and no Technical Specification change resulted.

DCN B0060E Replacement of Drywell Fire Damaged Cabling Unit 2 t

t'CN B0060E replaced sections of cables and conduits which were damaged during a drywell fire. The replacement sections of cable were spliced to undamaged portions where possible. These cables and conduits were returned to their pre-fire configuration as nearly as possible within the restrictions of current design criteria. No changes were made to the function of any cable, any circuit schematic, or the operating logic of any component/system. New junction boxes and supports were installed where required to accommodate the splices in the repaired cables. This modification had no affect on the consequences of any accident previously evaluated in the UFSAR. The changes to the raceway allowed installation to current design and installation standards such as those for cable fill and pull tension. This modification had no affect on the reliability of the modified equipment. Consequently, the probability, of a circuit malfunction was not deemed to be increased. No unreviewed safety question was created and no Technical Specification change resulted.

0 DCN M0102B Installation of Flange Clamp Leak Repair Assemblies on Regenerative Heat Exchangers Unit 3 DCN M0102B provided for use of flange clamp leak repair assemblies'or Unit 3 regenerative heat exchangers 3A and 3B. This modification minimized leakage from regenerative heat exchangers 3A and 3B and did not alter the design, function,-or operation of the RWCU system as described in the UFSAR.

Installation of the flange clamp leak repair assembly ensured all material and installation requirements were met. This- change has been evaluated to meet seismic Class II requirements and not endanger any safety related equipment.

The qualification of the involved heat exchangers was not affected. The possibilities for accidents presented by heat exchangers with repair clamps added are the same as those for heat exchangers without these changes. In general, the addition of a repair clamp would not affect the results of a malfunction of a heat exchanger. If the malfunction involved a leak at the flange joint where the repair clamp was installed, the clamp should act to minimize leakage reducing area contamination. Therefore, no unreviewed safety question was created and no Technical Specification change was required.

DCN H0140A SGTS Moisture Indicating Switches and Moisture Element Replacements Units 1, 2 and 3 This modification replaced and relocated non-safety related, seismic class II SGTS system moisture indicating switches and replaced safety related, seismic Class I (pressure boundary only) moisture elements, which provide a high humidity alarm for the associated filter train. The previous components are no longer manufactured and they require repairs to be made by the manufacturer. Since the design criteria and specifications for the components are satisfied, the system functions and operates in the same manner as previously evaluated in the UFSAR. This equipment has been seismically mounted and environmentally qualified and has been selected to operate and function in the same manner as the original installation. In addition, the modification did not adversely affect the function, qualification, or operation of any other plant system. No unreviewed safety question was created and no Technical Specification change resulted.

DCN W0209A Permanent Service Air Secondary Containment Penetration Unit 2 t n DCN W0209A provided a permanent SA connection which penetrates the secondary containment to support the ILRT program. This four inch SA line has valves 2-ISW-033-1190 and 2-ISW-033-1191 inside and outside of secondary containment, respectively. The SA line penetrations are located on the south reactor building wall at approximately elevation 570. To install this line, it was necessary to core drill a six inch hole in the 'reactor building wall. After installation of the SA line, the hole was covered by an anchor plate and sealed. No setpoints, process parameters or valve stroke times were affected by this proposed modification. The equipment function remained as presently described in the BFN UFSAR. The failure of this relatively small size penetration is well within the bounds of leakage already evaluated in the UFSAR. The radiological consequences of a failure of this secondary containment penetration seal along with other penetration seals has been evaluated; The Safety Evaluation for ECN E-2-P7104 evaluated the seal for increases in leakage and it was found to be bounded by an accident previously postulated in the UFSAR. A report by EgE Engineering evaluated the seal for seismic design and it was found to be acceptable. No unreviewed safety question was created and no Technical Specification change was required.

DCN W025ID, W0252D and W0296D Close Carbon Dioxide Pressure Relief Dampers Units 1, 2 and 3 The carbon dioxide pressure relief (mechanically actuated backdraft) dampers, located in the unit 1 and 2 computer room and all three auxiliary instrument rooms, were originally installed upside down. The improperly installed dampers do not reliably open at design pressure and do not automatically close. DCN W0251C sought to modify the backdraft dampers such that they would consistently open at the design pressure set point and automatically close.

The design modification, a counterweight/wire assembly, was coordinated with the damper vendor, Ruskin. However, during field testing, the dampers would not automatically close. Testing personnel had to manually close the dampers after the functional test. This modification was performed to disposition CARR BFP890565. As a resolution to this damper problem, a decision to lock the dampers closed was made. Other HVAC in the room will provide adequate ventilation with this small pathway sealed. Pinning the backdraft dampers shut and adding door sweeps did not create the possibility of an accident of a different type than previously evaluated in the UFSAR. Pinning the dampers shut'nd adding door sweeps will lessen the severity of damage to equipment in affected areas by permitting a more rapid and sustained C02 fire suppression concentration to be achieved. Inadvertent C02 release (with dsmpers pinned closed and door sweep installed) will not create structure over-pressure concerns. Therefore, the activity did not increase the consequences of a malfunction of equipment previously evaluated in the UFSAR. No unreviewed safety questions or Technical Specification 'change resulted.

ECN P0399 R6 480V Shutdown Boards and 120 VAC HC Buses Upgrades Units 1, 2 and 3 t v Nuclear Engineering analysis performed determined that the existing electrical design for the Class lE 480V Shutdown Boards 1A, 1B, 2A, 2B, 3A and 3B and the

't Unit 2 Class lE 120V AC 1&C buses did not provide adequate voltage regulation, circuit protection and breaker coordination. ECN P0399 provided the design necessary to adequately protect, coordinate and regulate voltage on these electrical systems.

ECN P0399 provided design to replace existing Class 1E, dual 15 KVA step down transformers between 480V Shutdown Boards lA, 1B, 2A, 2B, 3A and 3B and their associated Units 1, 2, and 3 IEC buses A and B with Class lE 75 KVA step down transformers. Also, Class 1E 50 KVA regulating transformers were installed in series with the 75 KVA transformers from Shutdown Boards 1A, 1B, 2A and 2B to Battery Boards 1 and 2 and panels 1-9-9 and 2-9-9. Design for 50 KVA regulating transformers from Shutdown Boards 3A and 3B to Battery Board 3 and panel 3-9-9 were incorporated in ECN E-2-P7161. The new/replacement transformers and circuits were located in their associated Shutdown Board rooms.

In addition to the above, this modification added or replaced cables, circuit breakers, fuses and disconnect switches, and modified existing circuit breakers in Unit 2 Battery Board 2, 480V Shutdown Boards 2A and 2B, and Distribution Panel 2-9-9. These changes were necessary based on the results of calculation ED-Q2000-87028 which has identified circuits which did not meet minimum voltage requirements due to inadequate cables sizes, and calculations ED-Q2000-870548, ED-Q2000-870549 and ED-Q2000-88086 which identified additional undersized cables and circuit breakers which required replacement or modification in order to provide proper terminal voltage, cable protection and breaker coordination. The proposed modification did not adversely affect the systems performance from that described in the UFSAR. The added Class 1E transformers, circuit breakers, cables, fuses and disconnect switches were fully qualified, designed in accordance with applicable criteria and provided adequate protection, regulation, and coordination of the affected auxiliary power systems. Any failure resulting in an undervoltage condition will cause automatic transfer to an alternate source. This modification was.made to ensure that the 120 VAC Instrument and Control Power Supply and 480V Shutdown Boards are properly protected, regulated and coordinated. This modification did not adversely impact the functions, operation and qualification of the affected systems or their associated systems and equipment.

This modification was evaluated and analyzed to ensure that no seismic or electrical separation impact existed. The proposed modification did not involve any initiation or failure not considered in the UFSAR. No unreviewed safety question was created and no Technical Specification change resulted.

DCN H0441A Qualified Air Flex Hose Replacement for MSIVs Unit 2 The as-installed 1/2 inch and 1 1/2 inch control air flex hose lines to the MSIVs (2-FCV-1-14, 2-FCV-1-15, 2-FCV-1-26, 2-FCV-1-27, 2-FCV-1-37, 2-FCV-1-38, 2-FCV-1-51, and 2-FCV-1-52) were found to be leaking and without proper installation and material documentation. This DCN replaced the existing 0

control air flex hose lines and fittings with new qualified flex hose .lines.

This modification also altered existing pipe ends, modified supports to accommodate the new flex hose lines, and seismically qualified the configuration. Control air accumulator supports were modified to meet seismic Class I requirements. This DCN also provided the material and documentation for the 1/2 inch and 1 1/2 inch control air flex hose lines to the MSIVs.

The replacement hoses were routed in proximity to where the existing flex hoses were routed. No new failures of equipment were introduced by the flex hose installation. Failure of a MSIV control air flex hose would result in the closing of the associated MSIV. UFSAR analysis of the transient caused by this failure is unaffected by this modification. The modification improved the reliability of the MSIVs with respect to 'the current leaking hoses. The design of the MSIVs was unaffected by this modification. The modification did not affect parameters which control the closing time of the MSIVs.

Neither the probability of the occurrence or the consequences of an accident or malfunction, nor the possibility for an accident or malfunction of a different type than previously evaluated in the UFSAR was increased or created. Therefore, no unreviewed safety question was created and no Technical Specification change was required.

DCN W0709A Installation of Maintenance Feeder Energizing Lights on 480V Reactor MOV Boards 2D and 2E Unit 2 t v The temporary emergency maintenance feeders were installed without any indication of bus energization. Thus, boards can have their normal and alternate feeders disconnected while the unannunciated maintenance feeder energizes the affected boards. This presented serious personnel risk. DCN W0709A installed Class 1E, fused local indicating lights in the temporary emergency feeder breaker compartments of 480V React'or MOV Boards 2D and 2E.

The Class 1E indicating lights illuminate while the Reactor MOV Board feeder cables are energized. Specifically, Compartment 5A of 480V Reactor MOV Board 2D (cable 2ES331-I) and Compartment 4A of 480V Reactor MOV Board 2E (cable 2ES3960-II) were modified by the installation of Class lE lights (1 each),

Class 1E fuses (2 each), Class lE fuse blocks (1 each) and Class 1E internal panel wire.

This modification had no adverse effect on the function or operation of the affected system or any other systems. Therefore, the capability of the affected system or any other systems to mitigate the consequences of the design basis accidents and anticipated operational transients for which they are designed was not affected. The credible failure modes associated with this modification are fuse open circuit without fault or overload.

Installation of fuses added potential failure modes which're enveloped by the existing design. This change complied with all design requirements for circuit protection and proper breaker coordination. Therefore. this modification did not add any new failure modes that could affect the function or operation of the involved systems or any other systems. No unreviewed safety question was created and no Technical Specification change resulted.

DCN W0782A Revised Design Calculation for Portions of RWCU Piping .

Units 1, 2 and 3 t n As identified in CAQR BFF880057, the RWCU piping between valve FCV 69-12 and the connection to RFW piping is designed for a temperature lower than its normal operational conditions. RWCU water exits the regenerative heat exchanger at 434'F under normal operating conditions, 448'F during hot standby conditions and approximately 545'F during blowdown. The current design temperature is 376'F. This DCN documented raising the design temperature, from 376'F to 545'F, for the RWCU system piping downstream of valve FCV 69-12 and included all interconnecting systems. DCN W0782 also documented moving the temperature interface with the RCIC system from valve FCV 71-39 to valve FCV 71-40. For the RCIC piping downstream of valve FCV 71-40, the design temperature was revised from 376'F to 545'F. For the piping between valve FCV 71-39 and valve FCV 71-40, the design temperature was revised from 376'F to 200'F.

A failure of a component or piping in the RWCU, RFW, RCIC or CRD which could either decrease the coolant flow to the reactor or release radioactive material in the primary or secondary containment has been evaluated in the UFSAR. Raising the design temperature of the RWCU and associated piping to meet operational conditions does not create any new type of malfunction or accident. Calculations prove the affected piping and components are fully qualified with the revised temperatures. The decrease in design temperature in the RCIC piping between FCV 71-39 and FCV 71-40 did not increase the probability of failure. Thus, no change in frequency class for potential pipe rupture resulted. Therefore, no unreviewed safety question was created and no Technical Specification change was required.

DCN W0898A Replacement of Properly Coordinated CBs in 480V Boards Unit 2 t'

This modification involved the replacement of Class 1E circuit breakers with new Class lE circuit breakers on 480V Reactor MOV Boards 2A, 2B, 2C and 3B, 480V Diesel Auxiliary Boards A and B, and 480V Control Bay Ventilation Boards A and B to provide proper circuit protection and breaker coordination. These changes were necessary based on the results of calculation ED-Q2000-87549 which has identified some 480V breakers may fail to protect cables served by these devices and/or may fail to operate properly due to motor acceleration time or inadequate breaker coordination. The only impact on the Auxiliary Power System caused by this modification is improved circuit protection and protective device coordination. All the electrical buses and loads downstream of the protective devices will function the same as they did before this modification. This modification has no adverse effect on the function, operation or qualification of the affected systems or any other systems.

Therefore, the capability of the affected systems or any other systems to mitigate the consequences of the design basis accidents and anticipated

operational transients for which they are designed is not affected. The credible failure modes associated with this modification are circuit breaker failure to close or circuit breaker failure to trip for fault or overload.

The replacing of circuit breakers with circuit breakers of different type or size to provide proper circuit protection and breaker coordination did not add or eliminate any failure modes. These changes complied with all design requirements for circuit protection and proper breaker coordination. No unreviewed safety question was created and no Technical Specification change resulted.

DCN W0899B Downsizing Radiation Monitor Breaker 302 From 40 AMPS to 30 AMPS Unit 3 t Ev t'his DCN changed unit 3, panel 9-9, cabinet 3, breaker 302 from TEF 40 AMP to TED 30 AMP to provide proper coordination with the upstream circuit breaker.

This breaker supplies power to RE-90-55, RE-90-57, and RE-90-58 in unit 3 reactor building, elevation 593, 565 and 519. The new 30 AMP breaker was installed to eliminate an equipment malfunction tripping of IKC Bus B due to a fault downstream of breaker 302. The downsized breaker also provided proper protection of downstream cables and wiring, while providing an adequate power supply for the three radiation monitors. If breaker 302 should spuriously if trip, or I&C bus B were deenergized, the radiation monitors will alarm for loss-of-power. No unreviewed safety question was created or Technical Specification change resulted.

DCN W0901C Breaker Replacement for 480V Diesel Auxiliary Board A Compartment 6D Units 1, 2 and 3 t ft v t Calculations ED-Q2000-87548 and ED-Q2000-87549 have identified 480-volt and 4KV circuit breakers that do not properly coordinate with other circuit breakers and/or do not provide proper protection for downstream electrical equipment. The trip device associated with 480V Diesel Auxiliary Board A Compartment 6D circuit breaker was an EC-1 circuit breaker. DCN W0901C was issued to replace the EC-1 trip device associated with this breaker with a PS-lA trip device. This modification conformed to all applicable design criteria, maintained all equipment seismic and environmental qualification, did not adversely impact the 10 CFR 50, Appendix R analysis, was based upon the results of approved electrical calculations, and eliminated existing circuit breaker coordination and inadequate circuit protection problems. The involved equipment required to ensure nuclear safety has a greater degree of reliability and availability, which helps ensure that acceptance limits for plant operation and radiological dose to the public are not exceeded. No unreviewed safety question was created and no Technical Specification change resulted.

ECN E-2-P0916 Installation of Various Power Supplies for PASS Component Unit 2 This partial ECN provided Class 1E Division I and Division II Instrumentation and Control AC Power from panel 9-9 through panels 9-54 and 9-55 to NEMA-4 termination boxes. This power will be used in a future ECN revision to power Class lE solenoid valves 2-FSV-043-40 and 2-FSV-043-42, Liquid and Gas Return; 2-FSV-043-50 and 2-FSV-043-56, RHR Liquid Sample; 2-FSV-043-57, Torus Gas Sample; 2-FSV-043-51, Drywell Gas Sample; and 2-FSV-043-76, RBCCW CW Supply. The new Class 1E circuit met seismic Category I and divisional separation requirements. The systems involved with this change will continue to operate within the design specifications and as described in the UFSAR. Proper circuit coordination was provided, thus the added cables will not propagate faults through its supply breaker/fuse. Fuses are not being installed (open circuit) for precaution since valves are to be installed later. All work involving penetration seals was completed in accordance with applicable Technical Specifications and design criteria. No unreviewed safety question was created and no Technical Specification change was required.

DCN W1073B Turbine Building CAM Replacement Units 1, 2 and 3 This modification replaced the existing CAMs 1-RM-90-250, 2-RM-90-250, and 3-RN-90-250 in the Reactor/Turbine Building Ventilation System; 1-RM-90-249, 2-RN-90-249, 3-RM-90-249, 1-RM-,90-251, 2-RN-90-251 and 3-RM-90-251 in the Turbine Building Exhaust System; and 0-RM-90-252 in the Radwaste Building Exhaust System, with more reliable microprocessor based Eberline CANs. The existing CAM chart recorders, in Panel 1-9-44 in the Unit 1 Control Room, were replaced with two Eberline control consoles with integral thermal printers.

In addition, the RRRMS interface was'eleted, and three report generator interface units with telephone modems which provide data links to the RADCON and Chemistry labs and the capability for a further data link to the Process Computer, were installed in RRRMS panel 3-9-58A, located in the Unit 3 computer room. Also, grab .sample points for the existing CAMs were removed, as the new Eberline CAMs have grab sample points with quick disconnects located on their carts. The new CAMs will monitor and record the various exhaust vents in the listed buildings for noble gas, iodine, and particulate activity.

The design for this modification had been prepared to ensure that all of the Radiation Monitoring System requirements continue to be met. Applicable Technical Specifications were not adversely affected by this modification.

System operation and responses were actually improved by replacing existing CAMs with more reliable microprocessor based Eberline CAMs. The components associated with this modification have been analyzed electronically and seismically to ensure that safety related portions of the RNS and other systems are not adversely affected. No unreviewed safety question was created and no Technical Specification change resulted.

~ '

DCN H1263B LRM Replacement of MS Radiation Monitors Unit 2 This modification involved the MS logarithmic radiation monitoring channels A, B, C and D and MS logarithmic radiation monitors 2-RM-90-136, 137, 138 and 139 located in the control room and mounted in power distribution panel 9-10.

Replacing MS logarithmic radiation monitors 2-RM-90-136, 137, 138, and 139 upgraded the system by installing new self-testing models, and adding reliability features of the new NVMAC line provided enhanced operation and reduced the potential for inadvertent plant ESF actuations. The new monitor electronic drawers were a "fit, form and function" replacement for the old ones and required no modification to existing plant wiring since they are plug

~

compatible. No modifications to the control panels were necessary since the new monitors fit in the same space and use the same drawer slides. The NUMAC monitors provided improved accuracy and resolution by reduction of the instrument drift rate via the incorporation of its zero and automatic gain features. In addition, the MS monitors are "fail safe" upon loss of power.

The modification did not adversely affect any systems function or operational characteristic, nor involved a change in the facility from that described in the UFSAR which could impact nuclear safety. Therefore, this modification did not involve an unreviewed safety question, or a change to the Technical Specifications.

DCN H1264B 10 CFR 50.49 Cable Replacements Unit 2

~ .

The DCN t v replaced numerous cables in their entirety with cables qualified to 10 CFR 50.49 requirements, revised cable terminations on numerous valve limit switches to give indication of valve in intermediate position, added environmental seals to numerous limit switches, and replaced damaged cables lES4446-II and 2ES4447-II terminated on FCV-084-019 and 020. This modification had no affect on the consequences of any accident previously evaluated in the UFSAR. No new pathways for the release of radiation or contamination to the general public or to plant personnel were created. This modification qualified the affected components to the requirements of 10 CFR 50.49. This environmental qualification ensures that the affected components continue to perform their safety function in an accident environment. The modification to the limit switches will allow the unit operator to determine if any of the affected valves are open or in an intermediate position and take action to prevent the release of contaminants to the environment. No setpoints, process parameters, or valve stroke times were affected by these changes. Equipment function remains as presently described in the BFN UFSAR with the exception of added valve position indication. Since equipment performance under accident conditions was enhanced and since the valve failure identification was improved by added position indication, the changes did not constitute a decrease in safety margin. Therefore, no unreviewed safety question was created and no Technical Specification change was required.

DCN W1290 Qualification of Cabling and Switches to 10 CFR 50.49 Unit 2 This modification replaced selected power cables in the radwaste system which had no affect on the probability of any previously evaluated accident. The circuit functional configurations were electrically equivalent to the unmodified circuit configuration since'he valve position indication revisions can not* affect valve operation (e.g., inadvertent opening) owing to the design of the control circuitry. The addition of conduit seals to the affected circuits left the circuits in a condition that is functionally identical to the unmodified circuits. The modification to the current configuration of the limit switches to give indication of valve intermediate position are the unit operator's indication of malfunctioning or stuck valves allowing maintenance or compensatory measures to address the faulty valve. The failure of any of the affected components would be no different than the failure of the existing components.

This modification also made changes to the Radwaste system by adding environmental seals to the conduits containing the cables for the position indicating lights, and revised limit switch terminations to provide indication of valve intermediate positions. Valves 2-FSV-077-002A, 2-FSV-077-002B, 2-FSV-077-015A, and 2-FSV-077-015B were affected.: The physical work on these valves had to comply with plant procedures to prevent the release of radioactive materials to the environment. No additional release paths were created for the release of radioactive materials to the environment and no additional exposure to radiation by the general public or .plant personnel was caused. The valves continue,to perform their function as before modification. No unreviewed safety question or Technical Specification change resulted.

DD-6-87-1325 (SEBFDD900063) Add Vent Valves to Two Drawings Units 2 and 3 This was a documentation change only resulting from,DD 6-87-1325 to revise appropriate flow diagrams to reflect the as-built plant configuration.

Specifically, vent valves 2-2-1578 and 3-2-1578 were added to flow diagrams 2-47E804-1 and 3-47E804-1, respectively. The valves are needed during initial system fill to vent any trapped air. This change documented the actual configuration and did not change the function or operation of the CDWS as, described in the UFSAR. The vent valve installation did not introduce any unanalyzed failure modes and 'did not adversely affect the safety functions of the CDWS. The vent line/valve installation for 2-2-1578 and 3-2-1578 affected the UFSAR in one respect only, which is that the vent is not shown on UFSAR Figure 11.9-la. No unreviewed safety question or Technical Specification change resulted.

0

'h 0

DCR 1676 Installation of a 1 Inch Drain Between Tube Bundle and Outlet Isolation Valve on Feedwater Heaters Unit 2

~ ~

t This one-inch drain connection was installed between the heater and the outlet isolation valve to allow draining the outlet side of the tube bundle and piping when heaters are isolated for maintenance. Presently, the outlet must be drained from inside the heater by removing a pipe plug or burning a hole in the divider baffle, which requires special dress to prevent personnel contamination from the water. Also, the drain would aid in handling leakage past the outlet isolation valve which hampers work inside the heater. Three credible failure modes were identified for these components following

~

installation in the plant. The first failure mode could be the rupture of the additional drain lines between the main condensate lines and the drain isolation valves, in which case condensate would be lost to the floor drain system. The water and steam exiting the line break could affect other condensate components and heater level and pressure control. A break could possibly cause a heater string isolation or isolation of that condensate line, which could lead to a feedwater isolation. Any effect on the unit from the loss of condensate would be bounded by the analysis on loss of feedwater.

Second, the drain isolation valves could leak (without a pipe rupture), in which case'the condensate would simply be returned to the condenser resulting in no adverse effects. Third, the drain isolation valve could fail to open which would be equivalent to a capped drain line negating the effect of this modification. Based on the above analysis, no unreviewed safety question was created and no Technical Specification change was required.

ECN L1965 Replacement of Buna-N Seals in Scram Solenoid Valves Unit 2 t'

t'he EQ Binder provided an analysis of the radiation effects on Buna-N which is the material of the original 0-rings used on the Scram Pilot Solenoid Valves (FSV-85 A 6 B). This analysis found that Buna-N will withstand the accident dose with acceptable margin. GE Service Information Letter, SIL 128 addressed a concern that cracking and deterioration of the Buna-N disc material was accelerated by long term exposure to the heat of the normally energized solenoid coil. As a result of this concern, GE-SIL 128 recommends that the valves should be rebuilt periodically to assure that the Buna-N Parts are not used in excess of seven (7) years. 'However, TVA EQ Binder QMDS allows replacement of Buna-N and all other non-metallic components every 5 years of service or 10 years from date of manufacture which ever occurs first.

GE, however still maintains that the total life (shelf and service) is seven years. This is inconsistent with test results and industry standard. TVA based the 10 years service life on the following:

Parker Seal's 0-Ring Handbook, under Age Control, states that MIL-STD 1523A establishes age requirements for a group of .specific Military Nitrile rubber specifications. MIL-STD 1523A states the shelf life of elastomeric 0-Rings to be 40 quarters (10 years). Parker further states that "field experience has demonstrated that storage condition, not time, determines the useful life of synthetic rubber seals."

Based on the above evaluation, this Safety Evaluation accepts the use of the original Buna-N 0-Rings to be sufficiently safe and reliable for Unit 2 Restart.

Hence, the as-installed configuration for Unit 2 is acceptable without implementing ECN L1965, revision 0. Use of the original 0-rings on the pilot scram valves did not adversely impact the scram function as described in the UFSAR. As determined in the EQ Binder the pilot scram valves are qualified to the Criteria of NUREG-0588 Category II or the DOR Guidelines of IE Bulletin number 79-01B (IEEE 323-1971). No unreviewed safety question or Technical Specification change resulted.

DCN 82544A Installation of Two Circuit Protectors in Series in Cables Traversing Redundant Raceways Units 1, 2 and 3 t v t This modification installed additional fuses in series with a previously installed breaker to allow mixing of non-safety circuits between redundant divisions of safety divisions. The change affected the Electrical Power System. The Electrical Power System provides power for the systems and components that require electrical power to accomplish nuclear safety functions as well as those systems and components that perform during normal operations. Failure of associated non-safety power cabling has the potential-for impacting both Division I and II circuitry when divisional separation is not maintained. However, a circuit breaker provides cable protection from postulated failures such as sustained overloads (e.g., high impedance faults) and phase to phase or three phase faults. The use of two circuit breakers in series or a breaker/fuse combination provides this protection for scenarios where the postulated single failure is the failure of a breaker to open.

Therefore, any potential failure of non-safety power cabling will not preclude safe shutdown and accident mitigation following design basis events. No unreviewed safety question or Technical Specification change resulted.

DCN H4310B Radiation Monitor Support Installation Units 1, 2 and 3 t

t'his DCN installed a support for radiation monitor 0-RE-90-130 including permanently installed lead shielding. The radiation monitor is located in the radwaste building and is installed in a non-safety related system. This liquid radiation monitor is provided to indicate when operational limits for the normal release of radioactive material to the environs are exceeded. The monitor will close valves to prevent release of liquid containing excessive radioactivity. In order to provide accurate readings by the radiation monitor, permanent lead shielding is provided to limit the amount of background radiation affecting the readings of the monitor. The design of the shielding is such that minimal removal of shielding is required for detector well cleaning and removal. The modification improved the structural integrity of the support for the radiation monitor and did not impact any safety related equipment. This modification did not adversely impact any of the functional requirements of the radiation monitor. The radiation monitor can not initiate any malfunction of any safety related equipment described in the UFSAR.

Therefore, no unreviewed safety question was created and no Technical Specification change was required.

DCN H4330A Adding Extra Gaskets to RHRSW Pump Units 1, 2 and 3 t' v t'CN H4330A provided for the installation of three gaskets in addition to an existing one between the pump base and the flange plate of RHRSW pump A2 in the intake pumping station. The DCN also modified existing supports, as necessary. The addition of these gaskets was an effort to reduce pump vibrations so that system reliability could be improved and associated maintenance with the existing pump configuration could be reduced. This modification did not conflict with the design criteria for the RHRSW system.

The function, performance, and qualification of the RHRSW system components was not adversely affected by the installation of these gaskets and support modifications. The RHRSW system and the intake pumping station as described in the UFSAR were not adversely affected by the addition of vibration reducing gaskets to RHRSW pump 2A. .Based on this, nuclear safety was not affected, an unreviewed safety question did not exist, and no Technical Specification changes were required.

DCN H4448A Replacement of EECW Check Valves Units.1, 2 and 3

~ ~

t Per 2-SI-3.2.3 Inspection, EECW system north and south header check valves O-CKV-067-502, -619, -622, and -671 have experienced excessive valve internal wear. The disc dog-ear/hinge pin holes have worn to the point that the disc drags the bottom (or side) of the pipe and can no longer provide the seal

necessary to prevent the backflow of water. DCN H4448 replaced these valves with Daniel Chexter 1601 type AC valves which have 316 stainless steel discs which are compatible with the type 316 stainless steel hinge pins. This change increased the reliability of the valves, and should provide an alternative to future disc replacement (e.g.,'he replacement material will allow for disc dog-ear/hinge pin holes to be welded and re-drilled in lieu of disc replacement). These replacement valves perform the same function and in the same manner as the existing valves. The replacement of the check valves did not alter the failure modes or performance of any system component or equipment: The check valves and associated piping and supports have been qualified for this application, and the seismic analyses and computations performed for this modification have demonstrated that the system and associated equipment remain qualified.

The EECW header check valves are not discussed in any of the requirements or associated bases of the -Technical Specifications, 3.5.C/4.5.C; however, the EECW system fqnctional requirements, equipment readiness mode, and valve position requirements are specified. This modification will not affect the function, qualification, or operation of the EECW system; therefore, the margin of safety is not reduced. No unreviewed safety question is involved and no Technical Specifications changes are required.

DCN H4514A Adding Portable Battery Chargers to Unit, Shutdown Board, and DG Batteries Units 1, 2 and 3 t v t Previously, due to the slow charging rate of individual cells, the entire battery bank and its associated DC System had to be maintained in the equalizing mode for extended periods. This time was minimized by single cell charging using a portable battery charger. The slow charging cells can be equalized separately, one at a time, while the remaining battery cells can be returned to float voltage.

This modification installed non-Class 1E, single phase, portable battery chargers (C & D Model ARR 2AC30) to facilitate single cell charging when required. Four portable battery chargers mounted on movable carts were procured. Also, seismically qualified mounting brackets were installed at predetermined locations in .each battery room to hold the carts in place during charging.

The DC output of a charger will be connected to a battery cell through Class 1E isolation fuses. These fuses are maintaining the required isolation between Class 1E batteries and non-Class 1E chargers'nd are housed in portable enclosures. The charger and the cable were procured non-Class 1E.

The alligator clips used to attach to a battery cell are insulated such that inadvertent shorting of the bus is not possible. 'he use of portable single cell charging units minimizes the time required to keep the battery in equalizing mode. The subject modification did not adversely affect the function, operation or qualification of the Class 1E batteries. With the

introduction of single cel,l charging, system components will be subjected to overvoltage for a lesser period of time resulting in their extended life.

Thus, this modification resulted in improved reliability of the plant battery system.

During single cell charging, the battery is not required to be taken out of service. Therefore, the operation of the 250V DC, 125V DC batteries and their associated systems and components was not adversely affected by this change given the credible failure modes. With the introduction of single cell charging,-the batteries will continue to perform their intended safety function as before. No unreviewed safety question was created and no Technical Specification'change resulted.

DCN W4917A Documentation Changes on Various Plant Drawings Units 1, 2 and 3

t v t As identified in CAQR BFP871153 Revision 1, various system piping design pressure and/or temperature discrepancies have been discovered on BFNP mechanical flow diagrams. The corrective action for the CAQR involved determining the maximum operational condition for each identified system interface and evaluating the interface piping and components for design adequacy. As documented in the evaluation calculation ND-Q2000-890008, changes were made to the appropriate system flow diagrams.

The piping and components subjected to increases in design temperature/

pressure requirements are fully qualified thermally and seismically, even considering code required margins. Thus, the failure modes of the subject piping and components are no different than currently exist. No modification changes were made to the existing physical plant configuration under this change. Existing piping and components have been shown to meet the revised design conditions for the condensate, auxiliary boiler, RWCU, RCIC, HPCI, RHR, CS, radwaste and CRD systems. Documentation of proper design pressures and temperatures did not cause any physical or operational changes to any plant system, structure, or component. No setpoint changes are involved.

Therefore, no unreviewed safety question was created and no Technical Specification changes were required.

DCN W5086 Documentation Change to Drawing 0-47E845-2 Units 1, 2 and 3 This change was a documentation change only to revise the appropriate flow drawing to reflect 'the as-built plant configuration and was associated with the resolution of DD 3-86-0054. This change consisted of adding valve 3-033-1525 and a one inch service air station at elevation 569 feet, R20-U in the Unit 3 Reactor Building. This discrepancy was identified during a walkdown.

This safety evaluation was required because UFSAR Figure 10.14-2b is affected.

There was no physical work associated with DCN W5086A. The addition of the subject components did not degrade the performance of the Service Air System.

When the service air station is needed for operational or maintenance activities, the Service Air Compressors would not be placed under any extreme load. Therefore, the performance of the Service Air System would not be affected. The addition of valve 3-033-1525 and its associated service air station'id not create a potential for creation of a new type of unanalyzed event. Complete loss of service air has been analyzed within the UFSAR. No unreviewed safety question or Technical Specification change was involved.

DCN W5237A Addition of Redundant CBs to Five Safety Related Cables Units 1, 2 and 3 t v t's a result of the BFN Cable Separation Discovery effort performed by NE/EE and the Ampacity Evaluation program, various changes to cables are required.

CAQR BFP870860 identified a number of non-safety related cables which have been routed such that they mix with both redundant Class 1E divisions. This is a violation of the separation requirements.

Exception to the separation requirements has been permitted by exception EX-BFN-50-728-1 and Procedure Method PM89-07 if non-safety related circuits are provided with a double means of Class 1E isolation. The double isolation may be provided by the addition of one set of Class 1E qualified protective devices in series with the existing Class 1E device. If the existing breaker is not qualified or adequate for the cable, double isolation may be provided by the addition of two sets of Class 1E protective devices, connected in series. Further, the protective devices used for double isolation must be coordinated with the upstream breakers.

This DCN provided double isolation for cables 2V1739, 3PL474, 3PL2275, 3PL2325 and 3R3005 by the addition of fuses in series with the existing breakers.

These additional or replacement double isolation Class 1E protective devices were selected and coordinated to ensure compliance with all applicable design criteria so that no new system interactions or events are credible which could affect nuclear safety. In addition, the affected equipment operates in the same manner and performs the same function as before. Thus, this modification did not alter the function, operation, nor qualification of any equipment.

The proposed modification did not add a new single failure or initiator which invalidated any previous UFSAR events or analysis. No unreviewed safety question was created and no Technical Specification change resulted.

4 DCN W5262A Minor Control Bay Chiller Modifications Units 1 and 2 t v t h This DCN replaced the four flexible connections in the RCW system line that serves Units 1 and 2 Control Bay Air Conditioning Water Chillers 1A and 1B.

The modification was required because the existing flexible connections and their installation did not conform to design requirements. In addition, this DCN documented the acceptability of the use of butterfly valves in lieu of gate valves for chiller lA and 1B valves 1-TCV-24-185 and -190, respectively.

The added/replaced flexible connections, and the penetrations, are passive components designed in accordance with applicable design criteria. This modification did not impact the function, qualification, or operation of the

~ EECW system, Control Bay Air Conditioning system, or any other system described in the UFSAR or the Technical Specifications. No Technical Specification change or unreviewed safety question resulted.

DCN W5328A Drawing Change Only Unit 3 t Ev t'his DCN W5328A was a documentation change only. The DCN revised drawing 3-47E852-1 to reflect the as-built plant configuration. As per as-built verification, the 2 inch drain pipe from the RHR Heat Exchanger Room, elevation 565, that ties into a four inch floor drain pipe in RHR Pump Room, elevation 519, does not exist and was deleted on the drawing. A 1 1/2 inch drain pipe from the sample station that ties into the existing 4 inch floor drain pipe was shown on the drawing to reflect the as-built condition.

Failure of the 1 1/2 inch drain pipe from the sample station connected to the Reactor Building Floor Drain Sump 3B would only result in a spill of radioactive waste inside the RHR pump room which would flow through, the floor drain back to the same pump. Therefore, a failure of this piping would be bounded by the current UFSAR analysis. This change is inside the Reactor Building and the original design connection was also inside the Reactor Building. This change did not affect the safety function or safe operation of the Liquid Radwaste System, or any other system as discussed in the UFSAR or in the basis for Technical .Specifications. Therefore, this change did not result in an unreviewed safety question or cause a revision to the Technical Specifications.

DCN W5337A Drawing Change Only Unit 2 t' v t'his DCN W5337A was a documentation change only. The DCN revised drawings 2-47E852-1 and 47W852-1 to reflect the as-built plant configuration. As per as-built verification, the 2 inch RHR catch pan drain pipe from the RHR Heat

, Exchanger Room, elevation 565, that ties into a 4-inch floor drain pipe, RHR Pump Room elevation 519, does not exist and was .deleted on the drawing. A 1 1/2 inch drain pipe from the sample station, that ties into the 4-inch floor drain pipe, exists and was shown on the drawing to reflect the as-built condition. Should a failure of the 1 1/2 inch drain pipe from the sample station containing radioactive effluent occur, the spilled liquid would be retained in the RHR Pump Room which would flow through the floor drain back to the same sump. Current UFSAR analysis covers potential failure of this drain line. Additionally, the deletion of the two inch drain from the RHR Heat Exchanger -Room does not increase accident consequences because this drain line is not taken credit for in this mitigation and two 4-inch drains still,exist.

The Reactor Building is designed to mitigate and contain this event.

Therefore, the offsite dose consequences of failure from this change are not increased. This change did not affect the safety function or safe operation

~

of the Liquid Radwaste System, or any other system as discussed in the UFSAR or in the basis for Technical Specifications. Therefore, this change did not result in an unreviewed safety question or require any changes to Technical Specifications.

DCN W5340A Drawing Change Only Units 1, 2 and 3 This change, DCN W5340A, was a documentation change only to revise appropriate flow diagrams and physical piping drawings to reflect the as built plant configuration. Current as-designed drawings show the service building trench floor drain tieing into the same drain header as the Radwaste Building laundry room floor drain line. This trench drain, as-installed, is actually connected to the drain header from the radiochemical lab shower/floor drain line. Both headers drain into the Radwaste building floor drain sump.

Failure of the piping connected to the Liquid Radwaste system would only result in a spill of radioactive waste inside of the Radwaste Building. The existing UFSAR failure analysis considers a complete loss of tank contents.

Therefore, a failure of this piping change would be bounded by the current UFSAR analysis. This change is inside the Radwaste Building and the original design connection was also inside the Radwaste Building. This change did not affect the safety function or safe operation of the Liquid Radwaste System, or any other system as discussed in the UFSAR or in the basis for Technical Specification. Therefore, this change did not result in an unreviewed safety question or require any Technical Specification changes.

DCN W5348A Addition of Check Valves in Service Air Lines to Offgas System Units 1, 2 and 3 v

t'CN W5348A documented the addition of service air to the Offgas system preheaters A and B. This change was initiated after a walkdown in which it was discovered that service air was piped to the Offgas system instead of 0

control air. The revised .version of this DCN, W5348B, added check valves in the one inch purge line from the service air piping to the preheaters to prevent backflow of offgas into the Service air system in response to 1E Information Notice number 79-08. This change is made to decrease the possibility of offgas releases to the environment in the event of a failure of the Service air system or misalignment of valves and loss of both service air compressors with the Offgas system in operation. In this event, contamination could spread to plant personnel through the use of the Service air system.

Under normal operation of the offgas preheaters, service air is isolated.

Purging'ith service air only occurs with the Offgas system shutdown. The addition of the 'check valves was a field and drawing change. This modification only affected nonsafety-related portions of the Service Air and Offgas systems which do not function to limit the consequences of equipment malfunctions. The Service Air and Offgas systems features and functions that limit the consequences of equipment malfunctions (e.g., the stack ducting) are not affected by this modification. No other plant system or feature that limits the consequences of an equipment malfunction was affected by this modification. Since the design, operation, and function of the plant systems were not changed by this modification, any events that might result from this change are bounded by the existing UFSAR accident analysis. No unreviewed safety question or Technical Specification change resulted.

DCN W5350A Drawing Change Only Unit 2 t v This change was a documentation change only to revise appropriate flow diagrams and physical piping drawings to reflect the as-built plant configuration by adding two 1/4 inch test connections on the Reactor Building Equipment drain sump discharge pipe downstream of check valves 77-637 and 77-640. Radwaste system components that serve as primary and secondary containment isolation were not affected because these drain lines are outside of the primary and secondary containment boundaries.

The affected Reactor Building equipment drain sump pump discharge piping is required to be Seismic Class II and the additional test connections have been analyzed by the 79-14 program. In addition, one test connection is on each sump pump discharge line. This change did not create any new common tie between the pumps. These normally closed test connections did not change the function or operation of the Reactor Building equipment drain sump pump system. The failure or inadvertent opening of these valves are not initiators of any UFSAR accidents. No credit is actually taken for accident mitigation from this system. The potential flooding from the added test connection is bounded by current UFSAR analyses and the spilled contents would drain back to the same pump. No unreviewed safety question was involved with no changes to Technical Specifications.

ECN P5399 Piping C02 Relief to Outside of Building Units 1, 2 and 3 t v change provided for a vent line from the pilot valves (FSV-39-35A and t'his FSV-39-35B) vent ports to the atmosphere outside the C02 Storage Tank Room.

This design prevented C02 from escaping inside the Diesel Generator Building whi.ch could create a hazard for personnel entering the area. The added vent line was not a pressure retaining part of the C02 System Master valve FCV-39-35. The added vent line did not create any new failures that would cause the inadvertent release of C02 from the Fire Protection System. The vent line was installed as a Class II feature which ensured that its failure during a seismi.c event would not adversely affect the ability of safety-related structures, systems or components from preventing and/or mitigating the consequences of an accident that would endanger public health and safety. No unreviewed safety question was created and no Technical Specification change resulted.

DCN W5713A Addition of Control Bay Water Chiller Isolation Handswi.tches-Unit 2 t'

This change implemented Special Requirement 28a.l of SEBFECN890015 for ECN E-2-P7215 for chiller B. ECN P7215 has relocated the power supply for chiller B from 480 volt shutdown board 1B to 480 volt shutdown board 2B which is a 1E division II source. Chiller A is supplied from 480V AC shutdown board lA which is a division I source. Separate handswitches are provided in each of the redundant chiller A and B circuits. Associated cabling is routed to maintain separation between the redundant control circuits for the two chillers. Therefore, single failure criteria are met. The installation is designed to meet the requirements for class 1E installati.on and calculations have been issued to assure the seismic integrity and operability of the chiller control circuits are maintained.

The load shed contacts in the circuit are required to trip the 480V AC loads to maintain the analysis of the function of the 480V AC system and the diesel generator standby AC power, system. The installation of separate qualified handswi.tches under manual operator control assures that a single failure could only affect one division of 480V AC power or one diesel generator. The installation is fully qualified by calculati:ons and analysis to be wi.thin the design basis of the UFSAR. The function of the HVAC system remained unchanged except the control was modified to allow compliance with Appendix R safe shutdown requirements. This modifi.cation did not create an unreviewed safety question or involve a Technical Speci.fication change.

DCN W5714A Replacement and Rerouting of Some Division I and II Cables-Uait 2 As a result of the BFN Cable Separation Discovery effort performed by NE and the Ampacity Evaluation Program, various cable changes are required including replacement, rerouting, and/or reassigning of division for divisional separation. This DCN replaced, rerouted, and reassigned divisional designations for selected cables. The cables that required modification as a result of the Cable Separation Program are identified. The cables that required modification as a result of the ampacity program have also been identified. The replacement of the motor connection box for 2-FCV-67-48 did

~

not have an adverse affect on the seismic qualification of the valve and the new connection box was seismically qualified for the application. The replacement starter coils for 2-FCV-70-47 had been selected based on the results of Calculation ED-Q2057-890065, Revision 0, which verified the adequacy of the new starter coils for use in this circuit. The starter coil installation was seismically qualified and had no adverse effect on the seismic qualification of either the starter or the RMOV board. The required cable, splice kits, terminal blocks and starter coils were procured Class IE and met their environmental requirements. All other required material was purchased with the appropriate qualifications.

The new cable conduit routing was very close to existing routing (location) and thus, the conclusions reached in the pipe break (Outside primary containment) analysis were not invalidated. The installation of the replacement cables did not affect the existing Appendix R analysis other than the relabeling of cable numbers. The replacement cables and the sections of existing cables utilized were selected or verified based on applicable design considerations, were routed in compliance with applicable design criteria, did not alter existing circuit design, and are in seismically supported raceways.

The splices made are qualified for their environment and installed per the vendor's instruction. The reassignment of divisional designations of cables had no effect on the cables or the components serviced by the cables. The modifications made to the components associated with the cables, terminal blocks and contactor coils, were in compliance with the applicable design criteria. Based on these considerations, this modification had no adverse impact on its associated systems and did not affect the function, qualification or operation of the associated end devices. Therefore, this modification did not involve an unreviewed safety question or require Technical Specification changes.

0 DCN H5918A Replacement of Pressure Transmitters on Radwaste System Units 1, 2 and 3 t'

This change consisted of replacing two existing obsolete liquid radwaste, differential pressure transmitters. Existing transmitter 0-PDT-77-44 (floor drain filter) cannot be calibrated to within its allowable tolerances.

Existing transmitter 0-PDT-77-83 (waste collector filter) required frequent calibration to maintain its allowable tolerances. Additionally, the measured process ranges were increased from 0-40 psid to 0-120 psid to allow use of new filter elements purchased by TVA Control 89NMA-74308A and tested in the system by Radwaste ST 89-6.

This change also consisted of replacing indicators 0-PDI-77-44 and 0-PDI-77-83 to reflect the new 0-120 psid range. The replacement indicators and transmitters utilized the original 10-50 mA range to avoid having to replace the existing loop power supplies and switches. A new differential pressure switch setpoint of 90 psid was required when the new Pall Filter elements were installed. Mhen the old Croll-Reynolds Filter elements are used as a backup, the pressure switches will be recalibrated to the old setpoint of 35 psid.

These transmitters provide annunciation for high filter differential pressure. Calculation ED-N0077-890284 verified that replacement differential pressure transmitters 0-PDT-77-44 and 0-PDT-77-83 will provide the accuracy and function required of instrument loops 0-PD-77-44 and 0-PD-77-83. The replacement instruments are more reliable than the existing obsolete and inaccurate instruments. This resulted in more reliable indication of high differential pressure annunciation signals on panel 25-17 for the floor drain and waste collector filters. This modification did not change any function or flow path of the Radwaste System.

The improved reliability and performance of these differential pressure loops ensures that the filters on the waste collector and floor drains will function properly. The design requirements for the Pall Rigimesh element are equal to or more stringent than the design requirements of the Croll-Reynolds elements. Use of the Pall Rigimesh elements for floor drain and waste collector filters cannot result in increased release of radioactive effluents to unrestricted areas. As .stated in UFSAR Section 9.2.6, a loss of tank contents within the Radwaste building would be contained within the Radwaste building which is designed to withstand a design basis earthquake. Failure of the filters would result in a contained volu'me of liquid in the waste sample tank or the floor drain sample tank ready for reprocessing.

Since the replacement transmitters and indicators have the same failure modes as the existing devices and pump failure modes have been shown to be bounded by UFSAR evaluation, all possible credible events associated with the change were bounded by existing failure analysis (UFSAR Section 9.2 and 10.16) and there was no potential for creatio'n of a new type .of unanalyzed event.

Further, no unreviewed safety question was created and no Technical Specification had to be revised.

DCN W6507A SGTS Train C Electrical Components 10 CFR 50.49 Qualified Units 1, 2 and 3

~ This modification replaced SGTS system Train C components to meet the environmental qualification requirements of 10 CFR 50.49. Limit (damper position) switches for O-FCV-65-051, 052, and 067 were replaced with similar switches produced by a different manufacturer. The new switches were tagged O-ZS-65-051A, 051B, 052A, 052B, 067A, and 067B. The power feed cables from JBl to the Relative Humidity Heater were replaced. Existing control cables from the limit switches to JBs,(routed through flex conduit) were replaced.

Power feed cables from the damper motors to JBs (routed through flex conduits) were replaced. This replaced equipment was selected and designed to ensure that the SGTS system will perform its intended safety functions'in the same manner as the original installation as described in the UFSAR. In addition, no other system was affected by this modification. No unreviewed safety 0 question was created and no Technical Specification changes resulted.

DCN W6795B Providing 10 CFR 50.49 Power Supply Transformer Cooling Fans Unit 2 to 480V Shutdown Board v t DCN W6795B was required because 480V Shutdown Board Transformers TS2A and TS2B must comply with 10 CFR 50.49 requirements and the transformer enclosure's internal cooling fans and their associated temperature switch, auxiliary control power transformer, fuse, fuse block, wiring, and terminal blocks were not 10 CFR 50.49 qualified. This DCN disabled the existing, internal cooling fan power supply and installed a new 240V lighting cabinet receiving power from existing 240V Lighting Board 2A which was used to provide power for the transformer cooling fans. The transformers current 1000/1333 KVA AA/FA nameplate rating was denoted on design drawings as being limited to 1000 KVA for all 10 CFR 50.49 engineering analysis and calculations. All cabling installed by this modification used conduit dedicated for this application.

The Safe Shutdown Analysis has incorporated conservatism by considering the affects of an additional, unrelated, unspecified fault in some active component or piece of equipment. This fault was assumed to result in the improper operation of a device which was intended to mitigate the consequences of 0

the accident. This fault was assumed for a range of effects such that there exists no single additional failure, of the type considered, that could worsen the computed radiological effects of the design basis accidents. Since the forced air cooling is not required, and the modification has been shown to eliminate a common-mode failure of Transformers TS2A and TS2B, this modification did not result in new equipment malfunctions which were not bounded by current UFSAR analysis. No Technical Specification change or unreviewed safety question resulted.

DCN W6811A Coordinating Circuits With Their Protective Devices for 480V RMOV, Shutdown, Vent Boards, and Diesel Auxiliary Boards Units 1, 2 and 3 In order to obtain proper breaker coordination, circuit breakers within the 480V AC auxiliary power system required changes to breaker settings, breaker types, and in some instances breaker trip devices per calculations ED-Q2000-87548 and ED-Q2000-87549.

480V RMOV Boards 1B, 2B, 1C, and 2E normal feeders and emergency feeders and 2C normal feeder breakers did not coordinate adequately with other breakers and required their STPU and their LTPU to be reset. Sensor rating SDB and trip devices for RMOV BD 2E breakers were also changed. Similarly, the normal and emergency feeder breakers of 480V RMOV Board 3B did not coordinate adequately with other breakers and required changes to their STPU, LTPU, and sensor tap settings.

The proposed modification was implemented in compliance with all applicable design criteria so that no new system interactions were credible which could affect nuclear safety. The affected components and equipment operated in the same manner and perform the same function as before the modification'. The failure modes of the replacement breakers (failed open and failed closed) were the same as for existing breakers. Furthermore, the modification did not alter the function, qualification or operation of any equipment within the affected systems or any other system. The changing of breakers, breaker settings and trip devices of the affected 480V RMOV, Diesel Auxiliary, Control Bay Vent and Shutdown Board did not involve an initiator or failure not considered in the UFSAR and did not increase the probability of an accident or involve a newly discovered accident, previously thought incredible to the point where it becomes credible. No unreviewed safety question was created and no Technical Specification change was required.

DCN W6819B Rerouting Cable 2M12, System 253 Unit 2 As a result of the BFN cable separation effort performed by NE, various cable changes were required including replacement, rerouting, and retagging. For the cable 2M12, it was determined by calculation that its existing size and configuration are inadequate due to excessive voltage drop. This DCN increased the cable size, rerouted, and retagged the cable. The intention of the previous design for ECN E-2-P7161, revision 0 was not achieved with the existing cable. The replacement cable had been selected based on applicable design considerations, was routed in compliance with applicable design criteria, did not alter existing functional circuit design, and is in seismically supported raceways. The new seals meet the requirements for flood protection and fire resistance, as applicable. This modification had no adverse impact on its associated systems and did not affect the function, qualification, or operation of the associated end devices. Therefore, this modification did not involve an unreviewed safety question and no Technical Specification change was required.

DCN W6821A Replace and Reroute a DG Protective Relay Power Supply Cable Unit 3 v

modification provided corrective action for CAQR BFP880199. Cable 3PL t'his 1915-Z3B incorrectly assigned a voltage level 4 in CAQR BFP880199 based on breaker setting. However, based on the full load current of the circuit, this cable should be assigned a voltage level 3. Also, previously this safety-related cable was identified (tagged) as a non-safety-related cable and routed in a non-safety related raceway. This DCN replaced the entire cable in new dedicated conduit and reassigned the cable to voltage level 3. This change brings the subject cable into compliance with the separation requirements presented in UFSAR Section 8.9 The replacement cable was selected based on applicable design considerations, was routed in compliance with applicable design criteria, did not alter existing functional circuit design, and was in a seismically supported raceway. The implementation of this modification did not require the breaching of secondary containment. The new penetration seals met the requirements for flood protection and fire resistance, as applicable. Based on these considerations, this modification had no adverse impact on its associated systems and did not affect the function, qualification, or operation of the associated end devices.

Therefore, this modification did not involve an unreviewed safety question or require Technical Specification changes.

DCN W6835A Replace and Reroute DG Protective Relay Power Supply Cable Unit 3 t'

This modification provided corrective action for CAQR BFP880199. The cable, 3PL1925 was incorrectly assigned a voltage level 4 based on breaker setting.

Based on the full load current of the circuit, this cable should have been assigned a voltage level 3. The replacement cable was selected based on applicable design considerations, was routed in compliance with applicable design criteria, did not alter existing functional circuit design, and was in seismically supported raceway. The new seals meet the requirements for flood protection and fire resistance, as applicable. Based on these considerations, this modification had no adverse impact on its associated systems and did not affect the function, qualification, or operation of the associated end devices. Therefore, this modification did not involve an unreviewed safety question or a change to Technical Specifications.

DCN W6839B Reroute of Drywell Cables Unit 2 t t v TVA NE performed an extensive evaluation to resolve problems associated with cable routing design and field installation practices for various categories of electrical cables installed at BFNP. The evaluation of BFNP cable separation problems is addressed in QIR EEBBFN88095 and CAQR BFP870860.

As a result of the BFN cable separation discovery effort, DCN W6839A was issued to replace, reroute and retag various cables inside and outside the unit 2 drywell associated with the drywell blower banks to resolve electrical separation problems. These non-Class IE power and control cables originate

~

from Class IE 480V Shutdown Boards 2A and 2B and Class IE 480V Reactor MOV Boards 2A and 2B through class lE isolation devices to the non-class 1E drywell blowers located inside the drywell. The installed configurations of these associated circuits were partially routed with cables of the opposite electrical division in violation of electrical separation criteria.

DCN W6839B revised the design issued under revision A to eliminate the rework of the cables inside the drywell by rerouting to proper division separation outside the drywell. The cables rerouted/replaced by this modification had no safety related function and their replacement had no affect on the function, qualification, or operation of any component or system required to ensure nuclear safety. The replacement and rerouting of the cables complied with all applicable design criteria. This modification had no impact on existing circuit functional design. In addition, the affected equipment operates in the same manner and performs the same function as before this modification.

No unreviewed safety question was created and no Technical Specification change resulted.

DCN W6842A Re-routing of Non-Safety Related Cables Units 1, 2 and 3 Cables 2NM1150, 2NM1160, 2NM1170, 2NM1180, and 2NM1190 failed to meet the separation requirements for a portion of their runs (within their cable trays which were inside the unit.2 drywell). Additionally, cables 2NM1150 and 2NM1160 failed to meet the'eparation requirements due to the source of their associated circuits (2MN1158 and 2NM1168) outside the drywell. Replacement cables were routed, inside the drywell, from penetration "EE" to their end devices using new and existing raceway. Cables 2NM1158 and 2NM1168 were re-terminated at'heir source panel in the correct division cabinet. Breaker number 238 for cable 2NM1158 in panel 9-9 cabinet 2 was relocated to panel 9-9 cabinet 3 breaker number 336.

Cable routing complies with the divisional separation requirements presented in BFNP UFSAR Section 8.9. Cables 2NM1150, 2NM1158, 2NM1160, 2NM1168, 2NM1170, 2NM1180 and 2NM1190 are not on the 10 CFR 50.49 list and do not require environmental qualification, but have been reviewed to ensure that their routing will not pass too close to any'ot pipe or to radioactive hot spots. Cables 2NM1150, 2NM1160, 2NM1170, 2NM1180, and 2NM1190 being replaced by this modification and cables 2NM1158 and 2NM1168 being relocated by this modification have no safety related functions and their replacement will have no adverse effect on the function, qualification or operation of any component, system or structure required to ensure nuclear safety. The subject cables comply with separation requirements presented in UFSAR Section 8.9.

Based on this, a change from one frequency class of occurrence of accidents evaluated in the UFSAR to a more frequent class did not result. No change to Technical Specifications were required and no unreviewed safety question was involved.

DCN W6844A 'hangeout of Breakers for.10 CFR 50, Appendix, R Unit 2 This safety evaluation was revised to reflect changes made to DCN W6844A via DCNs F9966A and F10665A. DCN F9966A replaced the originally proposed replacement breakers in 480V RMOV Boards 1A, 3A and 3B from TJK types to THJK types. The changes to DCN W6844A resulting from DCN F10665A are detailed below. For the other three 480V RMOV Boards the reevaluation showed that the existing breakers required replacement with TJK125 breakers. The THJK125 (as well as the originally proposed TJK125) breakers require larger compartments than those housing the original breakers. DCN F10665A located the replacement breakers in different compartments rather than using the existing compartments'. Because of unavailable appropriate hardware for these replacement breakers, the existing breakers were retained in their present location and connected in series with the new additional breakers.

This modification replaced existing breakers with adequate size Class lE breakers to provide adequate protection against cable auto-ignition as required by 10 CFR 50, Appendix R. The replacement breakers were sized and selected to ensure proper coordination with the upstream breakers, and to prevent the non-safety related cables from reaching a temperature which could degrade other cables installed in the same raceway. This modification did not adversely impact safety related circuits associated with the source panel.

Therefore, based on these considerations this modification did not involve an unreviewed safety question or require any Technical Specification changes.

DCN W6848 Reroute a Class lE Cable From Non-Divisional Raceway to Divisional Raceway Unit 2 As a result of the BFN Cable Separation Discovery effort performed by NE, various cable changes are required including replacement, rerouting and/or reassigning of division for divisional separation. This DCN replaced, rerouted,- and reassigned divisional designations for cable 3B180-Bl. This DCN did not involve modifications to the source nor the end device associated with the cable.

The cable was reassigned from Suffix lE to Division Bl. The cable was originally routed in non-divisional raceway. The replacement cable i.s now routed from the source to the end device in dedicated conduit. Cable routing complies with the division separation requirements presented in BFN UFSAR Section 8.9. The new cable was routed in raceway that is seismically supported. The new cable conduit routing was reviewed and the conclusions reached in the pipe break (outside primary containment) analysi.s are not invalidated. The installation of the replacement cable did not adversely affect the existing unit 2 Appendix R analysis other than the relabeling of cable number, correction of fire zones that cable traverses, and the deletion of requirements for manual action at 4kV Shutdown Board 3EA for a fire in fire zones 2-3. The conduit routing was reviewed for compliance with the normal control power associated circuit requirement and the results of the analysis in calculation ED-N0999-880700 were not invalidated. The functional design of the circuits associated with this cable was not affected by thi.s modification. The cable was on the 10 CFR 50.49 list, required environmental qualification, and was reviewed to ensure that its routing will not pass too close to any hot pipe or to any radiation source. No unreviewed safety question was created and no Technical Specification change resulted.

DCN 6852A --Replacement of Five System 31 Cables Units 1, 2 and 3 As a result of the Ampaci.ty Evaluation Program, vari.ous cable changes were required including replacement and rerouting. This DCN replaced and rerouted cables. This DCN did not involve modifications to the sources nor the end devices associated with these cables.

Cables ES3750II, ES3780-II, ES3788-II, ES3789-II, and,ES3825-II failed to meet the ampacity requirements for a portion of each cable. Replacement cables were routed from the source using existing and new raceway, including the cable tray in which the existing cable failed the ampacity requirements, to JB2646 (except for cable ES3750-II which was reduced from number two AWG to number six AWG at the splice and routed in the new raceway to its end device). The replacement cables were spliced to the existing cables at JB2646. The replacement cables and the sections of existing cables being utilized were 0

selected or verified based on applicable design criteria, did not alter existing circuit functional design, and were in seismically supported raceways. The splices made were environmentally qualified for their environment and installed per the vendor's instruction. This modification did not affect the function, qualification, or operation of the associated end devices. No unreviewed safety question was created and no Technical Specification change resulted.

DCN H6867A Relocating Flow Sensing Probe to Upstream of Offgas Dehumidified and Upgrading Flow Instrument Unit 2 ri ti t v t'he non class 1E 2-F-066-lllA/B instrument loops measure and record the Unit 2 off-gas system flow rates. The current instrumentation for gas flow measurement was provided by GE during plant construction. The existing offgas high and low flow instruments are obsolete and are difficult to calibrate.

'l This design change was recommended per GE SIL 264 which states that in many plants the sensors are installed in a process pipe area having a high moistur'e content. Both the existing and the new design utilize thermal dispersion type flow instruments. This sensor probe ascertains airflow by measuring the amount of current required to heat and maintain a probe at a set temperature.

Moisture particles impinging on the sensing element and evaporating can excessively cool the element, thus requiring more current to heat the probe than if condensation were not present. This results in a false indication of increased flow. The design relocated the flow elements to a piping section downstream of the dehumidifier and eliminated the moisture problem. This location was determined to be acceptable per Quality Information Request/Release (QIR BFEBFN89106 RO) and Calculation MD-N2066-900015.

A non Class lE Kurz Instruments Model 455 flow element, transmitter, and electronic package replaced the existing Rosemount flow elements, transmitters (2-FS/FT-66-lllA/B), and GEMAC flow indicating switches (2-FIS-66-lllA/B).

Local flow indicators (2-FI-66-lllA/B) replaced the existing flow indicating switches and were used in combination with the new probes. The low flow instrument setpoint for normal plant operation was changed from four SCFM to six SCFM to increase instrument accuracy at the low end of the 0-30 SCPM operating range. The existing non Class 1E GEMAC flow recorder was replaced with a non Class lE Leeds and Northrup Speedomax 100 flow recorder (2-FR-66-111) in Control Room Panel 2-9-53. New cables (2R3365 and 2R3362) were pulled to the relocated flow devices (2-FT-66-lllA/B) from the existing panel 2-25-95. Existing cables (2A125 and 2A126) were routed from the new 2-FI-66-lllA and 2-FI-66-lllB locations to panel 2-9-53. Implementation of this DCN did not introduce any new failure modes which could adversely affect the safety function or operation of any plant system. This DCN is acceptable from a nuclear safety standpoint. No unreviewed safety question or Technical Specification change resulted.

DCN H6910A Replacement of Reactor Building CANs Units 1, 2 and 3 t v I

This modification replaced 28 obsolete CANs located in various locations of all three units (and unit common) and also replaced the PCLD CAN (2-RM-90-256), for Unit 2 only, with more reliable Eberline CAMs. The existing CAN recorder and PCLD CAN chart recorder, located in panel 9-2 of each control room were replaced with an Eberline controller, console, and chart paper take up reel. The RRRMS interface was deleted as well as the external grab sample points, due to internal grab sample points being provided in the Eberline CANs. These CAMs monitor the noble gas, iodine, and particulate activity at their respective sample point. The flexible hoses on the sample supply and return lines for 2-RN-90-256 were replaced with one inch piping which utilizes threaded connections at the CAM. Additionally these "sample and return lines were seismically qualified to meet seismic category II/I requirements with up to 10 lbs/foot of lead shielding.

The replacement components added by this modification satisfy the requirements of the applicable design documents and have been seismically qualified. In addition, the components added have been evaluated and found to have no impact on the RMS except the improvement of instrumentation providing alarms locally and in the control room. The redesign of the 9-2 control panels only affected component location on their respective panels. Failure modes of the new equipment are no different than those of the previous equipment. This modification did not adversely affect other systems or components. This modification did not change setpoints, calibration intervals or function test intervals for any CAM. The 28 CAMs are not addressed in Technical Specification and no changes to the Technical Specifications were required and no unreviewed safety question was involved.

DCN W6983A Relocation of System 84 Components for 10 CFR 50.49 Unit 2 t v modification relocated components in the control loops for control valves t'his FCV-084-019 and 020. The functions of these flow control valves are to confine airborne radioactive materials to the primary containment to prevent an uncontrolled radioactive release to the environment and to vent gases from primary containment in order to help control the concentration of combustible gases inside primary containment following a LOCA. This modification relocated flow modifiers FM-84-19B, -20B (electric to,pneumatic signal converter) from a harsh to an essentially mild environment to meet the environmental qualification requirements of 10 CFR 50.49.

Solenoid valves FSV-84-19, and -20 were relocated to place them in the pneumatic signal line between the valve positioner and the valve actuator to comply with the corrective action for CARR BFP880640. The previous location of the valve positioner and FM between the valve actuator and the FSV restricted the venting of air from the actuator.

This modification added additional cables which are spliced to existing cables. It extended pneumatic signal lines and control air supply lines which added piping and fittings. Additional supports were added for the new conduits and piping. A failure of any of these new components could result in a loss of air pressure to control valve FCV-084-019 or FCV-084-020, thereby closing the valve. This will result in the failed valve being unavailable for venting of the primary containment post accident. This failure is acceptable as it is not a nuclear safety-related function. Since the failure action is to close the valve, the primary containment function will be maintained with an electrical or pneumatic failure. No Technical Specification change or unreviewed safety question resulted.

ECN E-2-P7020 Modify Condensate/Feedwater Sampling Equipment Unit 2 The modifications upgraded the in-line chemistry instrumentation in order to effectively monitor water chemistry; replaced or rerouted sample lines; and replaced sample pumps, the sample equipment, and instrumentation in the condensate/feedwater samples subsystem or the Sampling and Water Quality System. The Sampling and Water Quality System does not have a safety function. The changes required by this ECN are confined to the turbine building and do not contact any safety systems. Power is provided by a non-class lE bus. Therefore, no safety function was affected. No unreviewed safety question or Technical Specification change resulted.

ECN P7047 10 CFR 50.49 Cable Replacements Unit 2 This modification replaced Class 1E cables, splices and terminations which did not meet the environmental qualification requirements of 10 CFR 50.49. These splices and terminations were determined to be unqualified for the environmental conditions in their respective areas. In addition to replacing splices and terminations, the ECN replaced cable sections, made minor modifications to conduit systems and added splice boxes as needed to de-terminate existing cables. The conduit system modifications were necessary to facilitate the field work and allowed proper termination of splices. There were ten unrelated safety systems involved in the cable rework. The location of the work was the only reason various systems were grouped together under this one ECN. The function and operation of'he affected equipment remained unaltered by this modification. No logic changes were involved. The new cable and splices provided an environmentally and seismically qualified circuit for the affected components. Coordination of circuit protective devices remained unchanged. No unreviewed safety question or Technical Specification change resulted.

ECN E-O-P7070, Rl Installation of Stainless Piping in Radwaste System Units 1, 2 and 3 t v This modification replaced rubber hose used to transfer spent resin between the Spent Resin Waste Drain Tank, the Condensate and Waste Drain Tank, and the Condensate and Waste Phase Separators E and F with stainless steel pipe. Ball valves '(0-77-1404 and -1405) were added to isolate each phase separator and the existing gate valve (0-77-1052) was installed horizontally to prevent resin buildup. The modification provided additional storage for spent bead resin and allowed greater utilization of the Spent Resin Tank. In addition, check valve 0-77-1051 was installed in a horizontal position above the resin slurry line tee with the flush water line entering at the top of the tee. The temporary modification installed under a TACF was incorporated in Amendment 4 of the UFSAR. The current modification did not affect the function or operation of the Radwaste system as it installed permanent piping that facilitated the transfer of spent bead resins, as described in UFSAR Section 9.3.4.1. Also, the configuration changes decreased the potential for personnel radiation exposure by eliminating the dead leg and reorienting valve 0-77-1052 to preclude resin accumulation. The addition of equipment and piping as part of this modification did not affect the function or operation of any system or structure as discussed in the UFSAR. No unreviewed safety question was involved and no changes to the Technical Specifications were required.

ECN E-0-P7075 Installation of a Spent Resin/Sludge Sample Station Units 1, 2 and 3 This modification, originally implemented under TACF 0-85-037-077 Revision 1, installed a sampling station for the cleanup sludge recirculation line located in the Waste Packaging Room at elevation 565 of the Radwaste Building. The sampling station contains both automatic and manual capabilities for sampling spent resin/sludge. In addition, a line is routed from the Service Air System to the sampler control box for actuation'f the automatic sampler. A new line is routed from the'emineralized Water System to the sampling station for flushing and decontamination purposes in the event of a spill. An isolation valve is installed on the recirculation line on each side of the sampling box. A drain line with valve was installed'on the recirculation line on each side of the sampling box. A drain line with valve was installed on the glove box to handle sampling spills and decontamination activities. This line drains to a portable collection flask installed by the field.

This modification did not change the intent of the design to monitor radioactivity in the Filter-Demineralizer effluent liquid of the RWCU system as specified in UFSAR Section 10.17.3. A seismic analysis was not needed for this modification. However, a deadweight analysis of the piping was performed. The installed piping conforms to ASME Section III Class 3, Class F (for Radwaste System) per BFP8317, RO. Failure of this newly installed

sampling station will be no worse than the present design in regard to the release of radioactive material since this does not change the radiation source, leakage flow rates, or any other parameter'which could lead to an increase in radiation release. Spillage of radioactive liquid in the Radwaste building would continue to be contained (confined) in this building (ref.

VFSAR Sections 9.2.6 and 12.2.5.1). This change will not increase the radiological dose to the public. This proposed sample station relocation decreased exposure to plant personnel since the station was installed in a lower radiation zone. The rerouting of the CDWS line was done to the non-safety portions of the CDWS located in" the radwaste building where equipment is non-seismic. Likewise, the rerouted service air line is non-safety related. Hence, the modification of these two systems had no impact on nuclear safety. No unreviewed safety question was introduced and no Technical Specification required changing.

DCN E-3-P7113 Rl DG Air Dryers Units 1, 2 and 3 This DCN added air dryers and aftercoolers, that were upstream of the check valve in the non-safety related portion of the system. The piping modifications and electrical conduit supports associated with the addition of the air dryers and aftercoolers were seismic Category II qualified and had no impact on the safety-related (seismic Category I) portions of the DSAS or any other system. In addition, these air dryers will eliminate future problems associated with corrosion products in the starting air system, increasing long term reliability. The installation of the air dryers and aftercoolers did not affect the operability or function of the DSAS, and did not impact nuclear safety. No unreviewed safety question was created and no changes to the Technical Specifications resulted.

DCN H7142A HPFP Flow Element Replacement Units 1, 2 and 3 DCN H7142A replaced existing flow element 0-FE-26-135 with a new Annubar Model Number DMT-25-14-SCH-STD-HA2-CA2S-GM5-IMDS flow element. In addition, a threaded connection with an isolation valve (tee off of the existing process instrumentation line) for mounting a test gauge was provided on the discharge of each fire pump to allow local measuring of the pump discharge pressure.

Head acceptance criteria have demonstrated the need for more accurate and reliable instrumentation to measure the necessary parameters. Very little margin exists between the fire pump operating curves and the Technical Specification limit.

The replacement of flow element 0-FE-26-135 with a new model provided more accurate and reliable instrumentation related to the required fire pump performance of 2500 gpm at 300 feet total developed head. The added instrument taps with isolation valves to the fire pump discharge line provided

additional capabilities to determine fire pump performance by providing a connection to mount test gauges. This modification did not affect HPFP system design parameters as presented in BFN UFSAR Section 10.11 nor did it affect BW fire protection features.

The possible failure modes associated with this modification are as follows:

(A) flow element gives inaccurate indication of flow (B) failure of element (C) failure of tees or associated piping/valves Failure modes (A) and (B) are the same as for the existing flow element.

The tees and valves added are designed for system pressure, meet TVA piping Class M requirements and are installed in compliance with all applicable design codes, specifications and procedures. Failure mode (C) has the same result as failure of the existing process instrumentation line which is non-safety related. Thus, no new failure modes are introduced which could affect HPFP performance, or any safety system. The components addressed by this modification are non-safety related and outside the safety-related piping boundary. No unreviewed safety question was created and no Technical Specification changes resulted.

DCN 87147A Replacement of Stack Gas Sample Pumps Units 1, 2 and 3 DCN H7147A provided equivalent replacements for the non Class IE Stack Gas Sampling Vacuum Pumps 0-PMP-90-152A and 0-PMP-90-152B. The existing offline pumps which draw a continuous gas sample from the stack for radiation monitoring have proven to be unreliable from a maintenance standpoint. The replacement pumps are non Class IE Metal Bellows MB-602 pumps which were procured in accordance with Design Criteria BFN-50-7090.

The systems, structures, and components affected are the Radiation Monitoring System, pumps 0-PMP-90-152A and 152B and associated piping used to facilitate the installation of the new pumps in Panel 25-39.

The Stack Gas Radiation Monitoring System measures the radiation level of releases from the plant stack, making a permanent continuous record of the observed radiation levels. It also provides a means of obtaining samples for laboratory analysis. If observed levels exceed preset limits, the monitoring system initiates an annunciator alarm.

The pump replacement did not adversely affect system performance from that described in UFSAR Section 7.12.3. Stack Gas Radiation System operation is not essential for any accidents nor is its failure the initiator of any accidents. Therefore, this modification did not increase the probability of the occurrence of an accident previously evaluated in the UFSAR. In fact, with the improved reliability of the new pumps, the RMS is capable of better performing its intended function as described in the UFSAR. No unreviewed safety question or change to Technical Specifications resulted.

ECN-2-P7161, Rl Modifying Voltage Regulation to 480V Shutdown Board 1A, 1B, 3A and 3B and 120V,AC IKC Buses Units 1, 2 and 3 Based on previous analysis performed, it was determined that the existing electrical design for the Class 1E 480V Shutdown Boards 1A, 1B, 3A and 3B and the Class lE 120V AC IEC Buses did not provide adequate voltage regulation, circuit protection, and coordination. The subject ECN provided the design necessary to adequately protect, coordinate, and regulate voltage on the electrical systems. The possible failure modes associated with this modification consisted of failure of the regulating transformers added by this ECN to Unit 3, electrical faults, inadequate circuit protection, regulation, coordination, or the circuitry failing open. These failure modes will result in annunciation in the control room for abnormal voltage on Unit 1 and 3 panel 9-9 and automatic transfer to an alternate power feeder. The regulating transformers, breakers, cables, switches, and raceway modified by this ECN were seismically qualified by calculations and were installed in compliance with all applicable design codes, specifications, and procedures. It is verified that the electrical design of breakers and cables are acceptable from ampacity, voltage drop, and short circuit consideration. The electrical modifications were implemented in compliance with UFSAR Section 8.9 to assure satisfactory electrical separation/isolation and physical separation. The use of the installed maintenance bypass will be controlled administratively in accordance with Technical Specification requirements. The possible failure modes were enveloped by the existing analysis. No new failures were introduced which could adversely affect the function or performance of the 120 VAC Instrument and Control Power Supply, 480V Shutdown Boards, or any other safety related system. No Technical Specification change or unreviewed safety question resulted.

ECN E-2-P7194, RO Change of EECW Flow Valves Units 1, 2 and 3 This modification involved the redesign of flow control valves 2-FCV-067-050, 2-FCV-067-051, and 0-FCV-067-053 from hydraulic to pneumatic actuation. This change was necessary based. on CAQR BFP871058 which was written to document the repeated failure of these valves to perform their intended safety function of closing upon low EECW header pressure to ensure adequate flow to essential EECW loads. These failures were attributed to silt blockage in the actuator lines due to the use of raw river water for hydraulic actuation. These control valves are normally closed backup supply valves to the portion of the Raw Cooling Water System which supplies the RBCCW heat exchangers and Control Air Compressors. Changing the actuation for EECW valves 2-FCV-067-050, 2-FCV-067-051, and 0-FCV-067-053 from hydraulic to pneumatic did not degrade the function of these valves. Valve closing is the safety function of these valves and the valves are designed to fail in the closed position on loss of air or loss of power. This modification will improve the ability to perform their safety function. The solenoid valves added to the control air supply to these EECW valves are procured Class 1E and are seismically qualified. The pressure switches are procured Class 1E and are seismically qualified. This modification also included the addition of a relay, pushbutton switch, and lamp in panel 1-25-32; two relays, three switches, and three lights in panel 2-25-32; and one light and one switch in panel 3-25-32. The switches, lamps, pushbuttons, and relays are class lE and seismically qualified.

Cables being added in fire zones/areas were routed in either conduit sleeves or conduit that terminates in board rooms. New seals met the requirements for floor protection and fire resistance, as applicable for those penetrations which have been breached during implementation of this modification. A mechanical travel stop to limit the opening of Control Valve 2-FCV-67-50 was necessitated by Appendix R considerations. The existing valve was equipped with a stopping device to limit the diversion of EECW flow from essential loads following an Appendix R fire which disables the capability of this valves to close.

The redesign of the control system for the operation of EECW valves 2-FCV-067-050 and -051 and 0-FCV-067-053 and the installation of a travel stop on 2-FCV-067-050 affects EECW, the Control Air System, and the 120V AC Unit Preferred Power Supply System. Failure of these added devices and/or components a'e enveloped by existing accident analyses and are evaluated in the UFSAR. Thus, a change from one frequency class of occurrence of accidents to a more frequent class did not result from this modification. No change to Technical Specifications results and no unreviewed safety question was involved.

DCN E-O-P7195, Rl Connection of the 4160/480V Bladder Tank Substation Units 1, 2 and 3 This modification was to connect the 4160/480 volt bladder tank substation to the 4160 volt north loop line, which is fed from 4160 volt cooling tower switchgear D, panel 7. The normal power source for 4160 volt cooling tower switchgear D is 161KV/4160V cooling tower transformer 1, and the alternate power source is 161KV/4160V cooling tower transformer 2. The purpose of this activity was to provide an. adequate source of electrical power for the common maintenance building 480 volt distribution panels A and B, and for the diesel fire pump house 480 volt distribution panel. None of the loads supplied power from 4160 volt cooling tower switchgear C or D, the loop lines, or any downstream electrical equipment perform safety functions. Connecting the bladder tank substation to the 4160 volt north loop has no direct or indirect effect which could initiate an accident or introduce a new failure which could cause an accident. Therefore, this modification did not create a possibility for an accident of a different type than any evaluated previously in the UFSAR. Therefore, no unreviewed safety question or Technical Specification change resulted.

ECN E-O-P7197 Replace Bolts in MSIV Tunnel Blow Out Panels Unit 2 t v t ECN E-0-P7197 required that a number of broken explosive bolts and installed non-explosive bolts be replaced in the Unit 2 Reactor Building Main Steam Tunnel. The newer bolts have been found by testing to rupture at higher values than the existing bolts, and subsequently blowout panel failure will occur at some pressure in excess of 36 pounds per square foot as currently specified in UFSAR Sections 5.3.3.4 and 5.3.4.1. The existing explosive bolts on the blowout panels in the Unit 2 Main Steam Tunnel were designed to relieve at 70 pounds per bolt, while the replacement bolts have been tested and found to rupture at values between 132 pounds and 150 pounds per bolt. BFN UFSAR

~

Section 5.3.3.4 and 5.3.4.1 currently states that these blowout panels are designed to relieve at a pressure of 36 pounds per square foot. The use of higher strength explosive bolts necessarily resulted in the subject blowout panels relieving at some pressure in excess of this 36 pounds per square foot value.

NE subsequently documented in a calculation that the high strength explosive bolts allowed by ECN E-0-P7197 are acceptable, and that the failure pressure of the blowout panels will not exceed 90 pounds per square foot. The use of higher strength explosive bolts for the blowout panels in the main steam tunnels was reviewed by Nuclear Engineering. The results of their review indicated that a blowout panel failure pressure of 90 pounds per square foot was acceptable for all units, and that the peak pressure in the main steam valve vaults would not be increased by using replacement bolts with failure characteristics consistent with the 90 pounds per square foot blowout panel loading. In addition, NE also determined that the most severe temperature profile in the main steam valve vaults was due to a postulated break that did not fail the blowout panels. It can be concluded therefore, that neither the pressure nor the temperature requirements currently being used for equipment environmental qualification would be affected by increasing the relief pressure limit for the blowout panels to 90 pounds per square foot.

Based upon the above, the change made per ECN E-0-P7179 and the subsequent change to the UFSAR allowing a relief pressure of 90 pounds per square foot for the blowout panels, had no adverse effects on safety and did not involve an unreviewed safety question. No Technical Specification changes were required.

ECN P7214 Replacement of SGTS Heater Initiating Flow Switches Units 1, 2 and 3 t' t Ev t' To meet the environmental qualification requirements of 10 CFR 50.49, this modification replaced the existing SGTS flow switches with flow elements located on the SGTS ducts, in a harsh environment and flow switches located remotely in an essentially mild environment. The previous setpoint was

2000-4000 CFM decreasing.. A setpoint of 3000 CFM (9.19 FPS) was used .with better defined upper and lower limits to assure that the 2000-4000 CM limits are not exceeded. SGTS flow switches 0-FS-65-20A and 20B (train A),

0-FS-65-42A and 42B (train B), 0-FIS-65-70A and 70B (train C) were replaced with FCI thermal dispersion type flow switches. The previous installed trains A and B switches were McDonnell and Miller "paddle type" mechanical switches which sense flow by means of a paddle inserted into the air stream. When'he flow rate was sufficient to overcome a bias spring, the, paddle rotated and actuated an electrical switch. Failure modes were for the paddle and/or the electrical switch to stick in either the open or closed position.

The previous installed Train C switches were Dwyer Instruments Inc.

differential pressure sensing devices with electronic amplifier. The electronics were subject to failures that could result in either a fail-open or fail-close contact position.

The FCI switch is an electronic type. As such it also can fail with the electrical switch in either an open or closed position. Since both of the previous switches (McDonnell, Dwyer) could fail in the same mode as the replacement switches (FCI), no new failure mode was introduced.

This modification replaced flow switches in the SGTS which were not 10 CFR 50.49 qualified. This equipment has been seismically and environmentally qualified and has been selected to operate and function in the same manner as the original installation. This modification did not introduce any unanalyzed failure modes and did not adversely affect the safety function of the SGTS or any other plant systems. No unreviewed safety question was created and no Technical Specification change resulted.

ECN E-2-P7215 Change Power Supply to Control Bay Water Chillers IB and 3A-Units 1", 2 and 3 v

changed the power supply to Control Bay Water Chiller 1B and Control t'his ECN Bay Water Chiller 3A to 480V Shutdown Board 2B and 480V Shutdown Board 2A, respectively. The ECN also added a transfer switch on 4kV Shutdown Board D and routed a cable from this switch to the 250V DC control power transfer switch on 480V Shutdown Board 2B.

According to calculation ED-N0031-880692, for a fire in Unit 3 (fire zone 3) the power sources for control bay water chil1er 3A and control bay water chiller 3B are not available. Changing the power supply for chiller 3A from 480V Shutdown Board 3A to 480V Shutdown Board 2A made water chiller 3A available for a fire in fire zone 3. The previous source was a lE Division I source as is the new source; therefore, the safety function of the system was not reduced and the design criteria were still met. For a fire in unit 1 (fire zone 1) the power sources for control bay water chiller lA and control bay water chiller 1B are not available. Changing the power supply for Water Chiller 1B from 480V Shutdown Board 1B to 480V Shutdown Board 2B, made Water Chiller 1B available for a, fire in fire zone 1. The existing source ie a 1E Division II source and the new source is a lE Division II source. Therefore, the safety function of the system was not reduced and the design criteria were still met. Calculations show that the cables were properly sized to the loads. The new equipment associated with this modification was qualified.

The power supplies meet divisional separation requirements and single failure criteria. In addition, the new power supplies are as reliable as the existing sources. Therefore, no unreviewed safety question was created and no Technical Specification change was required.

ECN E-2-P7216 10 CFR 50.49 Temperature Switch Installation to RWCU Heat Exchanger Room Unit 2 t'

This modification removed from service existing Renewal temperature switches used to detect RWCU line breaks. They were replaced with environmentally qualified RTD and IEEE Class lE-qualified Rosemount ATU. The ATUs were installed in existing RPS Panels 2-9-83, 2-9-84, 2-9-85, and 2-9-86. The RTDs in the RWCU heat exchanger room and RWCU system pipe trench were located in the positions presently occupied by the Renewal switches. The RTDs mounted in the NS valve vault and the RWCU pump rooms 2A and 2B were wall mounted near the ceiling and away from any heat producing equipment. The RTD locations proposed in ECN P7216 were used in the environmental analysis calculation and were found to be acceptable. The purpose of this modification was to implement the corrective action of CAQR BFP890138 which specified that a 10 CFR 50.49 qualified leak detection system be designed, procured, and installed to detect and isolate RWCU pipe breaks including critical cracks in the MS valve vault, RWCU pump rooms and the RWCU heat exchanger room.

The new equipment is fully qualified for the new installation and performs the identical function as the original installation which is to provide a RWCU isolation valve closure signal. Since the new equipment will provide a faster and more reliable isolation signal, the modification will enhance the way in which the RWCU isolation function is implemented. Because of the instrument channel redundancy, electrical separation, and logic designed into the system, a single failure of any part of the system will not prevent isolation of the RWCU primary containment isolation valves nor is spurious operation more likely to occur.

The existing PCIS logic design remains unchanged. The margin of safety was enhanced by installing instruments that provide quicker response to a temperature rise indicative of a pipe break. Calculations have been performed to determine the analytical limits for the RTDs in each of the monitored areas and to determine the setpoints for the RTDs in each area. The type of design (e.g., analog loops) utilizing ATUs has been analyzed by NRC and found to be generically acceptable at BWR facilities. Therefore, no unreviewed safety question was created. Technical Specification change submittal 289 was processed for updating to this modification.

U ECN E-O-P7226 Change 480V Shutdown Board Tap Settings Units 1, 2 and 3 4

The existing transformer tap settings are based on calculations ED-Q2999-880575 R3 and ED-Q3999-890102 RO for fuel load. However, to support Unit 2 restart, calculation ED-Q2000-870026 R4 was prepared to determine the required tap settings for Unit 2 operation with Units 1 and 3 defueled. The tap settings were issued under ECN E-0-P7226 RO and were required prior to Unit 2 restart. These transformer tap settings will accommodate both the normal and accident loads under maximum and minimum voltage swings of the off-site power supply grids. The voltage range is 485kV to 545kV for the 500kV system and 165.5kv to 170kV for the 161kV system. In addition to these settings,,this ECN also changed the normal power supply for Control Bay Vent Boards A and B to 480V Shutdown Board 1A and 3B, respectively. The supply from 480V Common Boards will be retained as emergency feed to these boards in case the normal supply is not available for any reason such as maintenance.

The existing auto-transfer scheme between the two power supplies was also disabled and transfer to emergency power supply will be administratively controlled. The tap settings were optimized for all transformers which feed the safety-related boards required for Unit 2 operation. The tap settings were established after evaluating the loading changes due to various plant modification packages issued prior to January 6, 1990. These tap settings are essential for maintaining adequate voltage on safety-related buses needed during normal and accident conditions. A single tap setting for each transformer is valid for all modes of plant operation except for common station service transformers TCSS A and B. Two settings are required for these two transformers: one for cold shutdown or refueling mode (lightly loaded condition) and the other prior to startup (fully loaded condition). In order to prevent a severe overvoltage condition that could result due to extreme light load conditions tap settings of 1.000 were required. Therefore, no unreviewed safety question was created and no Technical Specification change was required."

ECN E-2-P7229, RO and ECN E-2-P7230, RO Installation of the SPDS Unit 2 t v This change installed Phase 1 of the SPDS for Unit 2 restart. The implementation of the Phase 1 SPDS was an NRC commitment per TVA letter to NRC dated December 19, 1989 (L44 891228 801). The principal purpose and function of the SPDS is to aid the control room personnel during abnormal and emergency conditions in determining the safety status of the plant and in assessing whether abnormal conditions warrant corrective action by operations to avoid a degraded core. The Phase 1 SPDS will provide information to Operations and Technical Support Center personnel. The Phase 1 SPDS user will be able to monitor the status of the following five critical safety functions'.

reactivity control, reactor core cooling and heat'emoval, reactor coolant system integrity, radioactivity control, and containment conditions.

-100-

A fault at the terminations in the safety-related panels is highly unlikely, but if one were to occur, the loss of the affected instrument signal loop would follow. However, the faults in the termination and wiring to the multiplexer that are likely to occur are the same as those failures that are likely to occur for existing cables and terminations. Based on the above discussion and the fact that acceptable electrical separation/isolation is provided per General Restart Criteria BFN-50-728 Revision 1, there is no violation to single failure criteria. Therefore, a failure as described does not create a condition which has not been previously evaluated.

Phase 1 SPDS input parameters are denoted in the Technical Specifications as protective instrumentation which initiate and control core and containment cooling systems (e.g., 2-PI-3-74A and 74B) and instrumentation needed to monitor the potential and actual breach of reactor coolant and containment integrity. Limiting Conditions for Operation as described by the minimum number of instrument channels per trip system are not physically impacted due to the insignificant increase in probability of instrument failures described above and by the single failure criteria conformance. Contingency actions are also described in the Technical Specifications in the event an instrument channel has failed (inoperable). 'The isolation of the data acquisition system prohibits external disturbances from affecting the power circuits serving SPDS. As a result, no unreviewed safety question or Technical Specification change is created.

ECN E-2-P7231, RO Install Cable for the SPDS Unit 2 t t v ECN E-2-P7231 installed cabling and wiring to connect PSPDS components. The PSPDS will partially implement the requirements of NVREG-0737 (Supplement 1) and will provide a central display of critical information. The information which will be incorporated into the final SPDS is not included on the PSPDS becaus'e it is not readily available to the operators .on the Unit 2 NCR panels.

The principal purpose and function of the PSPDS is to aid the control room personnel during all abnormal and emergency conditions in determining the safety status of the plant and in assessing whether abnormal conditions warrant corrective action by operators to avoid a degraded core. The PSPDS is a non-safety system and is not required to mitigate the consequences of any abnormal or emergency events. The addition of the PSPDS equipment did not impact the operation, function or response of systems or equipment. No unreviewed safety question or Technical Specification changes were created.

DCN W7300, W7301 and M7302 Reroute EECW Anti-Siphon Vent Piping Units,l, 2 and 3

~

t ~

t v The EECW anti-siphon vent piping air supply was rerouted from inside the secondary containment to the outside environment through the secondary containment wall. The air supply enters the vent piping by passing through a debris screen which is designed to prevent the entry of debris. This screen is designed to remain in place following a seismic, event and during high winds.

-101-

1 This modification was the result of CAQR BFP 880735 which identified a potentially radioactive release pathway to the outside environment caused by air being drawn from inside of the secondary containment through the EECW anti-siphon valves. The rerouted vent piping associated wi.th anti-siphon valves67-554 and 67-597 was seismically analyzed in stress problems Nl-267-4RC and Nl-267-4RD. The revised support calculations ensure compliance with seismic Class 1 criteria. These piping and support modifications did not alter the function of any structure, system or component. No unreviewed safety question was created and no Technical Speci:fication changes resulted.

DCN H7612A Component Upgrade to Transformer Yard HPFP Units 1, 2 and 3 DCN H7612A involved non-safety related HPFP system equipment which is located within the transformer yard. The DCN implemented the following modifications: repaired or replaced parts which are no longer available for the 4-inch Model D-1 Viking Deluge Valves; replaced train components on Maxitraol Model Sentry 3000 deluge valves; replaced non-UL/PM approved 3/4 and 1 inch female threaded spray nozzles* with UL and/or FM approved male threaded spray nozzles; and installed test tee connections on piping between deluge valve discharge pressure switches and,pressure switch isolation valves so that these switches could be tested locally via pressure simulation. The basis for Technical Specification 3.11 is to ensure the operabi.lity of fire protection systems for the purpose of providing adequate fire protection in portions of the facility where safety-related equipment is located. Equipment affected by this modification was not required by any Technical Specification and did not support any portion of fire protection systems that provide protection for safety related equipment. In addition, this modification did not change the function or operation of the HPFP or any other system analyzed in the UFSAR.

This modification did not adversely affect the ability of any plant system to perform required safety functions. Therefore, nuclear safety was not affected, an unreviewed safety question did not exist, and no Technical Specification revisions were required.

DCN W7636A Support Modifications to Portions of Radwaste System Unit 2

~

t ~

DCN W7636 modified, replaced and added supports for the Unit 2 radwaste system piping. This piping is part of the Unit 2 d'rywell floor drain sump pump di.scharge lines and associated valves which run from the drywell penetrations Z18 and X19 to approximately the reactor'uilding equipment drain sump. No piping was rerouted as a result of this modification. The added snubbers provide for the movement, of this piping due to thermal growth. The new supports and snubbers ensure that this piping will be capable of mitigating design events. All credible failure modes for this piping were bounded by the existing UFSAR analysis. As a result, no 'unreviewed safety question was created and no Technical Specification change was required.

-102-

4 DCN W7778 Replacement of Barksdale Pressure Switches Unit 2 Non-class 1E Barksdale switches 2-PS-68-93 and 2-PS-68-94 have an adjustable setpoint range from 50 to 1200 psig. This wide pressure range and the excessive drift associated with the existing switches causes unacceptable inaccuracy in the lower pressure ranges.

To resolve the problem of inadequate switch accuracy and excessive drift, these pressure switches were replaced with Class lE, Static-0-Ring pressure switches with a setpoint range of 20 to. 180 psig and an accuracy of 1.0 percent of upper range limit which gives a better accuracy and better stability near the 100 psig setpoint. The Static-0-Ring pressure switches, however, only contain a single microswitch (there are no Class lE pressure switches available on the market which contain two independently actuated microswitches). In addition, one of the two contacts from each pressure switch was removed from the current valve control logic for valves 2-FCV-74-53 and 2-FCV-74-67. This contact was logically redundant to other logic which controls these valves and is not required for proper operation of any logic required for Technical Specification compliance.

The margin of safety defined by the bases for Technical Specification 3.2.A/4.2.A (Primary Containment and Reactor Building Isolation Functions) and 3.2.B/4.2.B (Core and Containment Cooling Initiation and Control) was improved by this modification because of increased instrument accuracy and reduction of failure modes caused by the deletion of logically redundant contacts.

The replacement pressure switches are Class 1E and seismically qualified. The mounting bracket was designed and analyzed to retain the seismic Category I qualification of the associated local instrument racks. The safety-related operation, function, and response for components and systems involved was not adversely affected by this modification. Therefore, this modification is acceptable from a nuclear safety standpoint. No unreviewed safety question was created and Technical Specification submittal 287 was processed to NRC covering this modification.

DCN 87845 Removal of Unit 3 DG Air Start Relief Valves Unit 3 DCN H7845A removed four of the five relief valves originally provided on each of the Unit 3 DGs starting air banks and re-oriented the relief valve remaining on the middle air tank in each bank. The remaining relief valves were orientated in accordance with the manufacturer's original requirements (i.e., mounted vertically).

-103-

TVA 64 [05.9 65) (OP.WP.5.85)

UNITED STATES GOVERNMENT Memoranctum TENNESSEE VALLEYAUTHORITY On each tank with a relief valve being removed, the relief valve and a 1" x 3/4" hex head bushing were deleted from the piping configuration and a one inch square head plug was inserted to plug the line where the relief valve was attached. This weight removal did not,adversely affect the seismic response of the air line or tank.

Eliminating four out of five relief valves per bank reduced the probability of system failure due to relief valve malfunction. Re-installing the remaining valves in -an upright position enhanced their operation and complied with requirements specified by the vendor. Thus, equipment reliability was enhanced due to the reduced probability of equipment malfunction and nuclear safety was improved. This modification did not change the operational function, setpoint parameters, or operating procedures for the DG Starting Air

~

system. Single failure criteria was met and Seismic Class I quality was maintained. Therefore, this modification did not introduce any new equipment failure modes to the DG Starting Air system, and so did not create a possibility for a malfunction of a different type. No unreviewed safety question or Technical Specification change resulted.

DCNs H7869A(U-1), W9831A(U-2), and H7870A(U-3) Modification to RPS Circuit Protector Trip Setpoints Units 1, 2 and 3 Qgy t' t'he purpose of the RPS circuit protectors is to protect components supplied power from the RPS buses from damage or malfunction resulting from sustained UV, OV or UF conditions. The purpose of this modification is to reduce or eliminate spurious RPS circuit protector trips which could result in unnecessary reactor shutdowns and challenges to safety-related and important-to-safety equipment.

DCNs H7869A, W9831A, and H7870A affected the RPS circuit protectors downstream of the RPS MG sets and alternate power transformer feed. All excess conservatism was removed from the RPS circuit protector UV and OV trip setpoints, resulting in an increase in the OV trip setpoint and a decrease in the UV trip setpoint. The UF relay setpoint was changed from 58 HZ to 57 HZ.

The UF relay contacts were wired in series with the existing 3-second timer as the UV and OV relay contacts were previously. As a result, the UF, UV and OV relay trips are delayed three seconds by the timer. Also, the process computer is presently set to alarm if RPS bus voltage decreases to 108 volts or increases to 132 volts which is completely outside the voltage range allowed by the circuit protectors. This modification changed the process computer alarm setpoints and scan rate such that the process computer will produce an alarm warning the operators for voltage drift before the voltage reaches the upper or lower RPS circuit protector voltage trip setpoints.

-104-Buy U.S. Savings Bonds Regularly on the Payroll Savings Plan

TYA 64 (OS-9 651 <op-wp.5 65)

UNITED STATES GOVERNMENT P MemoTanctum TENNESSEE VALLEY AUTHORITY These changes were made in an effort to reduce the number of spurious RPS circuit protector trips. The new setpoints allowed increased voltage and/or frequency variations without the possibility of RPS component damage or malfunction with due consideration given to relay drift, deadband, and setting inaccuracy. The RPS will still perform its required safety function of protecting the reactor core and coolant pressure boundary as it did before this modification. Neither the setpoint changes nor delaying the underfrequency trip for three seconds had any effect on the response of the RPS to its. instrumentation signals because a channel trip will still function without any added delay as it did prior to this modification. Loss of RPS power because of a circuit protector trip will continue to result in a single-trip-system trip (half-scram) as .before this modification. Reducing the possibility of unnecessary reactor scrams and equipment challenges, while

~

continuing to adequately protect the most limiting, components supplied power from the RPS buses, increases the margin of safety. This resulted in no unreviewed safety questions. Technical Specification submittal 286 was provided to NRC covering this modification.

H7887A Disabling Fixed Spray Fire Protection Unit 2 DCN r v t'n Portions of the installed fixed spray system not required for fire suppression have been deleted from the Unit 2 design basis since they are not needed to satisfy any 10 CFR 50, Appendix R requirement. To prevent spurious operation of, the fixed spray system components and station fire pumps, the installed fire protection system was modified as follows: Panels 25-281A,25-288, 25-290,,25-291,25-316, 25-317,25-320, 25-292 and all associated interconnecting wiring, cable, sensors, detectors, relays, alarm modules, flow solenoid valves, flow control valves, and hand control valves associated with the fixed spray system in cable tray zones 2A, 2B, 2C, 2D, 2E, 2F, 2J, 2K, 2L, 2N, 2P, 2Q, and 2R in fire zones 2-3 and 2-4 were abandoned in place. Fire detector and alarm modules EA, EB, EC, ED, EE, EG, FA, FB, and FE located in fire zones 2-3 and 2-4 were disabled by installing jumpers to maintain a complete circuit to the alarm modules which remained energized after the modification. The annunciator windows at panels 9-8 and 9-20 from the above alarm modules were replaced with blank legend plates. All fire pump start circuits associated with the installed fixed spray system in fire zones 2-3 and 2-4 were disabled. Pipe caps or blind flanges were installed downstream of the isolation valves in the fixed spray lines supplying water to fire zones 2-3 and 2-4. The isolation valves were maintained in the open position to prevent overpressurization of the downstream piping.

The DCN did not alter any assumptions previously made in evaluating the radiological consequences of the accidents contained in the UFSAR. The abandoned fixed spray components are not required for fire mitigation. The modification did not adversely impact the operation, function, or qualification of the fire suppression and detection systems. This modification was evaluated to ensure that no seismic design or electrical separation impact exists. Therefore, no unreviewed safety question was created and no Technical Specification change was required.

-105-Buy U.S. Savin<~~> since these jumpers were only approximately two feet long, no new calculations were required.~~

Setting the TOL heaters to their maximum setting (115 percent) to eliminate the potential for spurious tripping of the equipment under elevated temperature condition did not adversely impact the function of the TOL heaters. Adding an interposing relay that requires less pick up voltage than the starter coil assured adequate margin for the control circuit operability.

The 480 Vac Control Bay Vent Board A uses molded case switches with Size 5 contactors. The molded case switches have demonstrated adequate continuous current carrying capability but the the Size 5 contactors were identified as potentially failing due to the elevated post accident temperatures.

Therefore, the Size 5 contactors were bypassed by a jumper cable across it by this DCN. Therefore, the limiting conditions of operation of the equipment and their operating time addressed in the Technical Specifications were not effected by this modification. No unreviewed safety question was created and no Technical Specification change resulted.

DCN W15741A Replacement of Bypass Valves on Control Room AHU Units 1 and 2 t v t DCN W15741A replaced existing normally closed 1 1/2 inch globe valves 1-31-522 and 1-31-526 with 1 1/2 inch gate type valves. These valves provide bypass flow around TCV-18 and -19, respectively. The TCVs are located on the chilled water return lines from Unit 1/2 Main Control Room AHU, and normally

~~-141-0571S~~

SAFETY EVALUATIONS FOR SPECIAL OPERATING CONDITIONS regulate flow through the AHUs. For a loss of control air to the TCVs, these valves fail in, the closed, position. This event requires the bypass valves to be manually opened to re-establish chilled water flow to the AHUs. The gate valves will ensure that the bypass line is capable of passing the required flow rate through the Control Room AHUs in the 'event of a loss of control air to the TCVs. Both valves are manual valves and preliminary testing has shown that the installation of the new gate valves have no effect on the remainder of the chilled water system. The replacement valves were manufactured to specifications equivalent to those for the original valves, and were qualified for the design conditions of the chilled water system. The credible failure .

modes for-the TCV bypass valves would be a failure of the chilled water pressure boundary integrity and the inability for these valves to be opened when required. The failure of one of these valves by either of the failure modes mentioned is mitigated by the fact that the Control Room AHUs are redundant to each other including the TCVs and their associated bypass valves. Therefore, it would require multiple passive mechanical failures in order to inhibit the ability of the Control Bay HVAC chilled water system to perform its required design basis functions. No unreviewed safety question was created and no Technical Specification change resulted.

DCN W15820A Replacement of Valves 1-31-530 and 534 to Increase Flow Units 1 and 2 t v t Two chilled water flow bypass valves which provide bypass around temperature control valves on the Control Bay Air Handling units were required to be replaced due to lack of required flow. The existing globe valves were replaced with gate valves to increase flow to 175 GPM. The 175 GPM is full chilled water flow required for full load requirements of the Unit 1 and 2 Control Bay Air Handling units. If the TCVs loose air supply to their controllers, they fail closed; therefore, a full flow bypass is required.

~~Recent testing discovered the existing globe valves were insufficient for the needed flow.~~

The replacement of globe type bypass valves with gate type valves cannot increase the probability of any accidents previously evaluated by the UFSAR, since the new valves will perform the same function as the existing valves and are qualified for the design requirements for the system. These valves are manually operated after a failure of control air to the TCVs for the Control AHUs. The replacement valves are manufactured to specifications

'ay equivalent to those for the original valves, and are qualified for the design conditions, of the chilled water system. The'refore, this valve replacement cannot increase the probability of a malfunction of equipment important to safety. No unreviewed safety question was created and no Technical Specification change resulted.

~~-142-0571S~~

~~k 1~~

~~0 0~~

~~1 1~~

( Fissions lh ( Iodines ( Particulates ( Tritium Gross ( Tri tium Dissolved ( Gross HONTH ( Activation ( Radioactivity ( Noble Gases ( Alpha I

~~January ND NO 1.16E-02 I 9.47E-03 2.27E-02 NO ND I~~

~~February NO ND 1.28E-02 I 7.04E-03 1.40E-02 ND 7.39E-05 (~~

~~I I Harch NO 1.48E-02 ( 8.11E-03 1.57E-02 ND 7.88E-05 (~~

~~I April ND ND ND 1.23E-02 ( 1.23E-02 1.56E-02 NO I~~

~~Hay NO NO 4.38E-05 1.80E-02 I 3.05E-02 2.69E-02 ND ND I~~

~~June NO 8 24E-03 I 2 '4E-02 1.78E-02 ND I~~

~~July 1.16E-02 I. 2.65E-02 2.17E-02 ND NO I~~

~~August ND 6.88E-05 2.65E-01 I 4.63E-02 1.28E-02 ND I~~

~~September ND 4.70E-05 4.44E-02 I 6.20E-02 1.09E-02 ND NO I~~

~~October ND 2.80E-05 3.60E-02 ( 2.13E-02 8.92E-03 ND ND I~~

~~November ND 2. 17E-07 5.22E-02 ( 2.54E-02 1.23E-02 ND I~~

~~NO NO ND is for non-detectable.~~

~~Variation in the data for gaseous releases have been correlated with the numbers of operating fans. There were no excursion of interest nor releases which exceeded Tech Spec limits.~~

~~-143-~~

~,

1989 OCCUPATIONAL EXPOSURE DATA NUMBER OF PERSONNEL (> 100 H-REH) TOTAL HAN-REH STATION UTILITY CONTRACT TOTAL STATION UTILITY CONTRACT TOTAL WORK 8 JOB FUNCTION EHPLOYEES EMPLOYEES AND OTHERS PERSONS EHPLOYEES EMPLOYEES AND OTHERS H-REHS RX OPERATIONS 4 SURVEILLANCE HAINTENANCE PERSONNEL 711 9 6 726 9.427 0. 135 0.041 9.603 OPERATION PERSONNEL 24 4 1 29 2.685 0.008 0.005 2.698 HEALTH PHYSICS PERSONNEL 96 11 0 107 37.927 0.067 0.000 37.994 SUPERVISORY PERSONNEL 21 0 2 23 0.250 0.000 0.023 0.273 ENGINEERING PERSONNEL 77 9 58 144 3.108 0.039 1.795 4.942 ROUTINE HAINTENANCE HAINTENANCE PERSONNEL 1317 10 44 1371 854.334 2.881 20.213 877.428 OPERATION PERSONNEL 23 1 1 25 0.556 0.000 0.097 0.653 HEALTH PHYSICS PERSONNEL 115 4 1 120 9.000 0.000 0.087 9.087 SUPERVISORY PERSONNEL 34 0 4 38 5.964 0.000 0.556 6.520 ENGINEERING PERSONNEL 116 12 101 229 16.712 1.602 33.062 51.376 IN-SERVICE INSPECTION MAINTENANCE PERSONNEL 17 17 0.601 0.000 0.000 0.601 OPERATION PERSONNEL 0 0 0.000 0.000 0.000 0.000 HEALTH PHYSICS PERSONNEL 02 22 0.069 0.000 0.000 0.069 SUPERVISORY PERSONNEL 0 0 0.000 0.000 0.000 0.000 ENGINEERING PERSONNEL 3 5 0.132 0.055 0.000 0.187 SPECIAL HAINTENANCE HAINTENANCE PERSONNEL 866 2 12 880 126.569 0.074 0.288 126.931 OPERATION PERSONNEL 2 0 0 2 0.025 0.000 0.000 0.025 HEALTH PHYSICS PERSONNEL 77 .1 0 78 1.313 0.015 0.000 1.328 SUPERVISORY PERSONNEL 27 0 1 28 1.614 0.000 0.030 1.644 ENGINEERING PERSONNEL 69 8 55 132 3.027 0.219 3.241 6.487

1989 OCCUPATIONAL EXPOSURE OATA NUM8ER OF PERSONNEL () 100 M-REM) TOTAL HAN-REM STATION UTILITY CONTRACT TOTAL STATION UTILITY CONTRACT TOTAL WORK 8 JOB FUNCTION EMPLOYEES EMPLOYEES AND OTHERS PERSONS EMPLOYEES EMPLOYEES AND OTHERS H-REHS I

WASTE PROCESSING MAINTENANCE PERSONNEL 0.000 0.000 0.000 0.000 OPERATION PERSONNEL 0.000 0.000 0.000 0.000 HEALTH PHYSICS PERSONNEL 0.088 0.000 0.000 0.088 SUPERVISORY PERSONNEL 0.000 0.000 0.000 0.000 ENGINEERING PERSONNEL 0.463 0.000 0 F 000 0.463 REFUELING MAINTENANCE PERSONNEL 29 29 0.214 0.000 0.000 0.214 OPERATION PERSONNEL 7 7 0.279 0.000 0.000 0.279 HEALTH PHYSICS PERSONNEL 16 16 0.202 0.000 0.000 0.202 SUPERVISORY PERSONNEL 0 0 0.000 0.000 0.000 0.000 ENGINEERING PERSONNEL 3 3 0.003 0.000 0.000 0.003 TOTAL BY JOB FUNCTION MAINTENANCE PERSONNEL 2940 21 62 3023 991. 145 3.090 20.542 1014.777 OPERATION PERSONNEL 56 5 2 63 3.545 0.008 0.102 3.655 HEALTH PHYSICS PERSONNEL 331 16 1 348 48.599 0.082 0.087 48.768 SUPERVISORY PERSONNEL 82 0 7 89 7.828 0.000 0.609 8.437 ENGINEERING PERSONNEL 276 31 214 521 23.445 1.915 38.098 63.458 Grand Tota1 3685 73 286 4044 1074.562 5.095 59.438 1139.095

~~CHALLENGES TO OR FAILURES OF MAIN STEAM RELIEF VALVES g~t None None g~t None All three units were shutdown during the entire reporting period.~~

~~-146-0577S~~

~~REACTOR VESSEL FATIGUE USAGE EVALUATION The cumulative usage factors for the reactor vessel are as follows:~~

Shell at water line 0.00620 0.00492 0.00431 Feedwater nozzle 0.29782 0.21329 0.16139 Closure studs 0.24204 0.17629 0.14360

~~-147-0577S~~

~~TRANSMISSION LINE COORIDOR HERBICIDE USAGE In 1990, no herbicides were used in the maintenance of the transmission lines rights-of-way for BFN.~~

~~-148-0577S~~

ML18033B648

Text

SUMMARY

Navigation menu

Search