ML18033B029
| ML18033B029 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 11/03/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML18033B028 | List: |
| References | |
| NUDOCS 8911130009 | |
| Download: ML18033B029 (23) | |
Text
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 ENCLOSURE SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPLEMENT TO THE SAFETY EVALUATION ON POST-FIRE SAFE SHUTDOWN BROWNS FERRY NUCLEAR PLANT UNITS 1
2 AND 3 DOCKET NOS. 50-259 50-260 AND 50-296
- 1. 0 INTRODUCTION The staff issued a Safety Evaluation (SE) on Post-Fire Safe Shutdown Systems for Browns Ferry, Units 1, 2
& 3 in December 1988 (S. Black to 0. Kingsley, December 8, 1988).
Since the issuance of the SE, the Tennessee Valley Authority (TVA and the licensee) has revised the engineering evaluations which support the 10 CFR 50 Appendix R submittals upon which the staff SE was based.
The licensee has also submitted written comments regarding the staff's SE identifying changes which should be made to incorporate the new information.
In addition, the staff has completed two Inspection Reports 50-259/89-13, 50-260/89-13 and 50-296/89-13 dated August 1, 1989 and 50-259/89-28, 50-260/
89-28, and 50-296/89-28 dated September 15.
1989 to address Appendix R post-fire safe shutdown of Unit 2 with both Units 1 and 3 shutdown and defueled.
This supplemental SE has been prepared to address the revised analysis for Unit 2 operation only and to incorporate the inspection findings for Unit 2 operation.
This supplemental SE considers the use of certain equipment in Units 1 and 3 for the shutdown of Unit 2 and assumes that Units 1 and 3 are shutdown with no irradiated fuel in their cores.
New information which has been evaluated by the staff for this supplemental SE consists of (1) Volume I of the Engineering Evaluations in suppo'rt of the licensee's 10 CFR Part 50 Appendix R submittal for Browns Ferry Nuclear Power Plant (BFN) (Gridley to USNRC, dated February 14, 1989),
(2) comments on the staff SE on Post-fire Safe Shutdown and the licensee's Appendix R Exemption Request (Gridley to USNRC, dated March 10, 1989),
and (3) various drawings, calculations and evaluations which were reviewed as part of the staff's two inspections listed above.
The Appendix R exemptions were issued fn the staff's letter dated October 21, 1988.
2.0 SAFE SHUTDOWN PERFORMANCE AND GOALS 2.1 Safe Shutdown The means of shutting down the reactor was described in Section 2.1 of the December 1988 SE and was addressed in TVA CoIIIIent 3 dated March 10, 1989.
In accordance with the discussion in Section 2.6 of this supplement, the RHR pump seal coolers are removed from the list of minimum post-fire safe shutdown 8911130009 891103 PDR ADOCK 05000260 P
PNV
systems (SSDS);
Protection of components required for safe shutdown is dis-cussed in Section 2.1 of Inspection Report 50-259/89-13, 50-260/89-13, and 50-296/89-13 (BFN IR 89-13).
Two inspection follow-up items were found in this area.
These items will be closed by a final NRC inspection before the restart of Unit 2.
2.2 Reactivit Control Reactivity control was addressed in Section 2.4.5 of the staff SE and Section 4.3. 1 of Vol. I of the licensee's Appendix R Engineering Evaluations.
There were no significant changes needed in the staff SE as a result of our review of the licensee's submittals.
Control of reactivity is discussed in Section 2.2 of BFN IR 89-13.
There were no adverse inspection findings.
2.3 Reactor Coolant Makeu Reactor coolant makeup was addressed in Sections
- 2. 1, 2.4.1, 2.4.6, and 2.4.8 of the staff SE; TVA comments 7 and 28 (TVA submittal, March 10, 1989);
and Sections 4.3.3, 4.4.1.2, 4.5.1, 4.5.4, 6.6, 6.7. 6.8 and 9.4.1 of Volume I of the Appendix R Engineering Evaluations.
As a result of TVA Comment 7, clarification is made that the condensate storage tank is the source of water for high pressure coolant injection (HPCI) for all cases of operation except for a fire in the Turbine Building where spurious closure of valve FCV-2-162 may require automatic transfer of pump suction to the torus.
Analysis by the licensee also determined that there would be adequate available net positive suction head (NPSH) on the residual heat removal (RHR) pumps to prevent cavitation even wi th the drywell coolers in operation.
In'addition, the drywell coolers do not need to be stripped for high impedance fault con-siderations (refer to Section 6. 1 of this supplement).
- Hence, the staff's position on the drywell coolers in Section 2.4.6 of the staff SE is changed and it is acceptable for the drywell coolers to operate in the event of high temperatures in the drywell.
Reactor coolant makeup is discussed in Sec-tion 2.3 of BFN IR 89-13.
There were no adverse inspection findings.
2.4 Removal of Deca Heat Removal of decay heat was addressed in Sections
- 2. 1, 2.4.1, and staff's SE and Sections 4.3.4, 4.4.1.3, 4.5.3, 6.2.5 and 6.6 of Appendix R Engineering Evaluations.
There were no TVA comments addressing this topic.
The staff concludes that the SE remains area.
Removal of decay heat is discussed in Section 2.4 of BFN There were no adverse inspection findings.
2.4.8 of the Volume I of the on the SE valid in this IR 89-13.
2.5 Process Monitorin Process monitoring was addressed in Sections
- 2. 1 and 2.4.9 of the staff SE; TVA comnents 2 and 29; and Sections 4.4.4, 4.5.11, and 6.3 of Volume I of the Appendix R Engineering Evaluations.
The staff has revised its position in regard to portable instrumentation for indication of suppression pool level and temperature.
The licensee has com-mitted to provide instrumentation in the control room and at the backup control panel for all units for direct indications of suppression pool level and temperature prior to the start>>up of each unit.
Hence, portable instrumen-tation will no longer be required.
Process monitoring is discussed in Section 2.5-of BFH IR 89-13.
There were no adverse inspection findings.
2.6 Su ort S stems and E uf ment Support Systems and Equipment are addressed in Section 2.4.2 of the SE, TVA comnents 8, 9, 10, 11, 12 and 13; and Sections 4.4.3.7, 6.2.1, 6.2.5 and 6.4 of Volume I of the Appendix R Engineering Evaluations.
Section 6.2.5 of the Volume I of the Appendix R Engineering Evaluations states that the RHR pump seals have been re-evaluated and their temperature ratings have been upgraded from 160'F to above 215'F.
The maximum suppression pool temperature has been determined to be less than 200'F.
Hence, there is no need for RHR pump seal coolers during post-fire safe shutdown operations.
The licensee has added the Units 1 and 2 environmental coolers and the RHR room coolers to the list of Safe Shutdown System (SSDS) equipment and components.
In Section 2.4.2 of the SE, the staff stated that only three diesel generators (OG's) are required for SSDS equipment and components.
This was incorrect.
The staff reviewed alignment diagrams for the Auxiliary Power System which were based on Unit 2 operation only.
The alignment diagrams are contained in Appendix J of Volume III of the Appendix R Engineering Evaluations which was submitted for audit.
The review showed that each of the eight diesels i,s required for a fire in at least one of the fire areas.
The number of diesels required to be in operation for a fire in any one fire area varies from three to five.
Support systems and equipment are discussed in section 2.6 of BFN IR 89-13.
There was one inspection follow-.up item in regard to inclusion of HVAG require-ments into post shutdown procedures.
This item will be closed before the restart of Unit 2.
2.7
~fl 1
0 Manual operations were addressed in SE Section 2.4.3; TVA conrnents 14, 15, 16, 17, 18, 19, 20, 21, and 22; and 4.4.3.7, 6.2.4, 6.1.7.2, 9.0 and Table 9-1 of Volume I of the Appendix R Engineering Evaluations.
r
In the staff SE, the staff addressed the actions to be performed and the local station at which actions would have to be taken for a fire in fire area 16.
The following is a list of changes to these actions and locations:
a.
The Unit 3 DG panels (25-270 A5B) are deleted from the location list because these are for the Unit 3 DG building HVAC system and the Unit 3 DG's will not be required for a fire in fire area 16.
b.
Valve positions will not be indicated at local panels.
This is allowable because the licensee will physically place all valve controls in the proper "RUN" position, forcing the valve to either go to or remain in the correct position.
c.
Item 3 in the SE in Section 2.4.3 (Manual Operations) should read:
"Ensure HPCI shutdown (in the fire affected unit) to prevent water intrusion into the main steam lines (10 minutes)."
The'time is not being changed.
d.
Item 4 in the SE in Section 2.4.3, "Trip 250V DC control power to 4kV and 480V SDBDS......", is deleted.
e.
Item 7 in the SE in Section 2.4.3 should read:
"Verify AC power to an EECW pump (20 minutes)."
The time is reduced from 30 minutes to 20 minutes.
f.
Item 8 in the SE in Section 2.4.3 should read:
"Initiate diesel auxiliary systems to support long term DG operation by aligning power to diesel auxiliary boards (30 minutes).
Also connect 125V DC distribution boards if needed (two hours)."
The 125V DC distribution boards are not needed for at least two hours because of the battery capacity of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
g.
Item 9 in the SE in Section 2.4.3 should read:
"Trip nonessential loads from DGs (20 minutes)."
The time is reduced from 30 minutes to 20 minutes.
h.
Item 16 in the SE in Section 2.4.3 should read:
"Close applicable steam line drain and reactor water clean-up (RMCU) branch connections to maintain condenser and radwaste system high/
low pressure interface valves to mitigate their potential spurious operations (600 minutes)."
The time is changed from 480 minutes to 600 minutes in accordance with Section 6. 1.7;2 of Vol. I of the Appendix R Engineering Evaluation.
Item 17 in the SE in Section 2.4.3 should read:
"Close a valve to prevent CST water overfilling the torus (fire affected unit) (600 minutes)."
The time is changed from 480 minutes to 600 minutes in accordance with an analysis contained in the ONE Calculation entitled "Appendix R - Required Mechanical Equipment for a Unit 2 Safe Shutdown".
These changes to the staff SE do not affect the staff's conclusions in Section 2.4.3 of the staff SE that the licensee's proposed manual operations are adequate subject to verification during the Appendix R site inspections.
Manual actions are discussed in sections 2.7, 4. 1 and 4.2 of BFN IR 89-13.
There were four inspection follow-up items associated with these sections.
These included emergency lighting (which has been corrected), clarification of safe shutdown instructions (SSI's),
and tagging of valves.
These items will be closed before the restart of Unit 2.
2.8 Modifications Modifications are addressed in Section 2.4.4 of the staff SE; TVA comments 23, 24, 25 and 26; and Chapter 8 of Volume I of the Appendix R Engineering Evalua-tions.
In regard to re-routing of cables for the main steam relief valves (MSRV's),
the licensee proposes to re-route the cables for three of the MSRV's in order to ensure that three MSRV's will always be available for performing manual depressurization from the control room (not necessarily the three that were rerouted).
In addition, all NSRV's can -be manually operated from the control room.
All six automatfc depressurization system (ADS) valves on all three units have manual control capability at the backup control panel.
The proposed modifica-tion will eliminate the connection to the backup control panel for 3 of the MSRV's on each unit.
For Units 1 and 2, two of the three MSRV's are ADS valves.
For Unit 3 one of the valves is an ADS valve.
Thus.
on Units 1 and 2, four of the six ADS valves will maintain manual control capability on the backup control panel.
On Unit 3, five of the six ADS valves will maintain manual control capability on the backup control panel.
The remaining MSRV's (nine on Units I and 2 and eight on Unit 3) will have manual trip isolation capability either on. the backup control panel or on the 250 V RMOV Board B (for the three relocated MSRY's).
In regard to cable modifications for the RHR system, the licensee has listed proposed modifications to be completed before restart in Section 8.3.2 of Volume I of the Appendix R Engineering Evaluations.
These were reviewed and determined to be-acceptable.
In regard to the drywell control air systems (DCA) and containment air dilution (CAD) systems, the modifications are to the CAD system so that a manual connection can be proviaed to the DCA system.
This was not clear in the staff SE.
The staff concludes that the proposed modifications are acceptable.
3.0 COMPLIANCE WITH 10 CFR 50 APPENDIX R SECTION III.G.2 For some of the fire areas associated directly with Unit 2 and cewen areas with Units I and 3, the preferred method of protection for Unit 2 operation is protection in accordance with Section III.G.2 of 10 CFR 50 Appendix R.
This requires dividing the safety related plant areas into fire areas and fire zones.
For the III.G.2 areas (2 and 25), the licensee has stated redundant shutdown trains for Unit 2 operation have been protected by one of the following methods:
a.
Separation of redundant trains by a 3-hour rated fire barrier; b.
Separation of redundant trains by a horizontal distance of 20 ft. plus automatic detection and suppression in the fire area; and c.
Enclosure of one redundant train in a fire barrier having a one hour rating plus automatic detection and suppression in the fire area.
The fire area and zone designations were based on the following information:
a.
Suppression system coverage; b.
Detection system coverage; c.
Fire area boundary and fire barrier ratings; d.
Fire door ratings; e.
Portable extinguisher locations; f.
Hose station coverage; g.
Heating, ventilation, and afr conditioning (HVAC) duct locations; h.
Fire damper ratings;
- i. Inventory of combustible materials;
Pipe/conduit penetrations in fire barriers.
In regard to combustible loadings, conservative values of heat release rates assuming total combustion of all fixed combustibles in each fire area and zone was utilized to maximize the effect of potential fire hazards.
Cable insul-ation was evaluated assuming cable tray sections are filled to the maximum design limits.
The entire reactor building for each unit is considered as a fire area.
This fs because the Reactor Buildings have large equipment hatches and open stair-ways on all floor levels.
Thus, the only solid boundary barriers are the walls separating the Reactor Buildings from each other and the walls separating the shutdown board rooms.
Host of the fire area boundaries are reinforced concrete floors, ceiling, and walls, concrete block walls, and a meta1 partition with gypsum wall board which forms a boundary for fire areas 17 and 18.
The hour ly ratings of reinforced concrete walls, floors and ceilings of the various areas were determined by comparing the thickness of reinforced concrete required for specific hourly ratings as determined from common references.
The licensee determined that a 6 1/2 inch thick reinforced concrete wall would be equivalent to a three hour rated wall; whereas, floors and ceilings of the same rating must be at least 4 1/2 inches thick.
Concrete block walls were rated by comparing the b1ock composition, shape and reinforcement with that of rated walls from the U. L. Fire Resistance Directory. Hollow block concrete walls were assigned a resistance rating of one and one-half hours.
A metal partition with gypsum wall board was assigned a fire resistance of one hour based on comparison with similar configurations in the U. L. Fire Resist-ance Directory.
Exterior walls which are designated as fire area boundaries were not required to be rated fire barriers if all of the following criteria are met:
a.
Combustible loading within the imnedfate vicinity of the wall is low.
b.
The wall is not required to separate SSDS inside the fire area from the redundant equipment located within the iomediate vicinity outside the fire area.
c.
The wall does not separate safety-related areas from non-safety-related areas that present a significant fire threat.
d.
The wall has not been designated as a rated barrier in previous fire hazard analysis.
The fire areas in the Reactor Buildings were further divided into fire zones which meet the separation requirements of Appendix R section III.G.2.
Where floor to floor separation is required between zones, water curtains are installed around equipment hatches and stairways.
The fire zones were ini-tially established to ensure that the redundant, trains of the RHR system are adequately separated.
The availability of the Auxiliary Power System was then determined for the zone.
After the zones were determined such that adequate protection is provided to the RHR and Auxiliary Power System, redundant trains of other required systems were protected by plant modifications.
In some
- cases, exemptions were requested by the licensee and approved by the staff [S.
- Black, (NRC) to S. A. White, (TVA) dated October 21, 1988j.
The following areas in the Reactor Building are provided with automatic suppression and detection systems:
a.
Partial coverage is provided on Elevation (EL) 519 for the HPCI turbine.
b.
Nearly complete coverage will be provided on EL 565 and 593.'.
Automatic sprinklers have been provided for protection of in situ combus-tibles or over important equipment. on EL 621 of the Reactor Building.
d.
An aqueous film forming foam system (AFFF) is provided for coverage on EL 639 for the reactor recirculation pump NG-sets.
e.
Water curtains have been installed around stairway and/or equipment hatches in the Unit 2 Reactor Building to provide floor<<to-floor separation for ELs 565, 593, and 621, where needed for fire zone separation.
f.
Water curtains have been installed at the ceiling of the RHR pump room at EL 565 to provide separation between the RHR pump room and RHR heat exchanger room.
Water curtains have also been provided over the door openings of the RHR heat exchanger enclosures.
Some of the suppression systems are in place from the original plant design and the "Plan for EvaluaCfnn, Repair, and Return to Service of Browns Ferry Units I and 2 (March 22, 1975 Fire)" (Fire Recovery Plan) requirements.
- However, many of the modifications are for the purpose of meeting Appendix R separation requirements.
The existing detection system was installed as part of the original plant design or as modification resulting from the Fire Recovery Plan which followed the March 1975 fire.
Both ionization and heat detectors are provided in selected plant areas.
Ionization detectors are located in:
a.
Sprinkler and areas of the Reactor Buildings on EL's 565, 593, and 621 and along the north wall on EL 639.
b.
The Control Building or EL's 593, 606 and 617 except for stairwells, corridors, and lunchrooms.
c.
Relay cabinets on EL 617 of the control building.
d.
On EL 550 of the intake pumping station, partial coverage in the North Bay.
e.
Cable tunnel between intake pumping station and Turbine Building.
f.
Diesel Generator Buildings Units I and 2; pipe and electric tunnel, and diesel auxiliary board rooms.
g.
Diesel Generator Building Unit 3, pipe and electric tunnel.
h.
Shutdown Board Rooms 3EA, 3EB,
- 3EC, and 3ED on elevations 565 and 583.
4KV Bus Tie Board Room.
j.
Turbine Building, EL 586 open areas and EL 565 cable tunnel.
k.
Battery Board Rooms on EL 593.
Heat detectors are provided in the cable spreading room, the Reactor Buildings and the Diesel Generator Building.
All areas containing fixed water spray systems have continuous strip thermal detectors in selected cable trays.
In the Reactor Building heat detectors are installed in the recirculation pump motor generator (HG) set area, the reactor core isolation cooling system (RCIC) turbine area, and the HPCI tur bine area.
In the Diesel Generator Building, heat detectors are used in the diesel genera-tor rooms and the oil pump transfer rooms to actuate the automatic carbon dioxide system.
Additional detectors are being added as par t of the Appendix R compliance program for Unit 2 prior to restart.
As part of this program, detectors are being added to the following locations:
a.
Heat detectors will be added in the Reactor Building RHR pump rooms ac the ceiling above EL 541 (Unit 2 ECN P0885) b.
Heat detectors will be added in the Reactor Building below the ceilings on EL's 565 and 593 in unsprinklered areas of (2-1, 2-3, and 2-4).
10 Heat detectors will be added with automatic suppression below the ceiling on EL 621 in Fire Zone 2-5.
(Unit 2 ECN P0885).
c.
Smoke detectors in the Auxiliary Instrument Rooms I, 2 and 3 on EL 593 of the Control Building will be relocated and additional detectors will be installed in Rooms 2 and 3
(ECN P0671).
The licensee requested by letter dated November 21, 1986 and was granted two exemptions from the requirements of Section III.G.2 of Appendix R (S. Black to S. A. White, dated October 21, 1988).
One exemption was for lack of 20 ft separation in the RHR Pump Rooms and Heat Exchanger Rooms.
This exemption was granted based on the low fire loadings, the automatic detection system and the water curtain in the RHR pump rooms.
The second exemption was for lack of 20 ft separation between redundant trains of safe shutdown components with no intervening combustibles.
This exemption was generic to the Reactor Building.
This exemption request was granted based on the placement of supplemental sprinkler coverage where ther e were intervening combustible loads.
The licensee also submitted engineering evaluations per Generic letter 86-10.
These evaluations include use of non three-hour rated fire door s in the reactor buildings fire area boundaries, lack of automatic suppression and detection on the refuel floor, lack of three-hour rated HVAC dampers in the reactor building boundaries, and the use of draft stops and water curtains around openings in lieu of one hour or three hour barriers.
All of these evaluations were reviewed by the staff and found to be acceptable.
The areas discussed in Section 3.0 of this supplemental SE were inspecred during July 17-21, 1989 and discussed in Inspection Report 50-259/89-28, 50-260/89-28 and 50-296/89-28.
The only adverse finding was in regard to the composite metal wall and its acceptability to the staff as a fire area boundary.
In the letter dated October 31, 1989, TVA committed to modify this wall prior to Unit 2 restart.
Other unfinished modifications are listed as inspection follow-up items in the inspection report.
These modifications will be completed before restar t.
4.0 ALTERNATIVE OR DEDICATED SHUTDOWN
'his is new.information which supplements the discussion in Section 2.4 of the staff SE.
The only fire area for which alternative shutdown is provided is Fire Area 16 which contains the control= room and cable spreading room for all three units.
No areas were identified as requiring dedicated reactor cooling or depressurizatfon beyond that associated with the minimum SSDS.
For a fire in Fire Area 16, if the safe shutdown instruction (SSI) is executed, the fire area will be abandoned and remote shutdown panel 25-32 will be manned along with other locations.
11 The cable spreading room is protected in accordance with the requirements of Section III.G.3 of Appendix.,R by an automatic detection
- system, an automatic preaction sprinkler system, and a manually activated C02 system.
The licensee also requested and.was granted two exemptions from the require-ments of Section III.G.3 of Appendix R in regard to Fire Area 16 (letter of October 21, 1988).
The first exemption was for lack of a fixed fire suppres-sion system in the main control rooms.
The exemption was granted based on the existence of an automatic detection
- system, 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> manning by operators, and three-hour fire rated adjacent barriers.
It was the staff's conclusion that the installation of a fixed suppression system would not significantly incr ease the level of fire protection in the main control room.
The other exemption was for the lack of detection and suppression in some parts of the control building.
This exemption was granted on the basis of low fire loading, and the availabi 1-ity of detection fn most areas with the exception of the stairways, corridors, lunch and computer services room.
Compliance with the fire protection requirements for alternative shutdown areas was inspected during the July 17-21, 1989 inspection.
The results of the inspection are presented in Inspection Report BFN IR 89-28.
There were no adverse findings in regard to fire protection of alternative shutdown areas.
- 5. 0 ASSOC IATEO CIRCUITS 5.1 Common Bus Concern Common bus concerns are addressed in SE Sections 2.4. 14. 1 and 2.4. 14.3; TVA Cogent 30; and Section 5.6 of Volume I of the Appendix R Engineering Evalua-tions.
The SE stated that the licensee proposed to strip all non-safe shutdown loads from the common bus in a timely manner as part of post-shutdown proce-dures to prevent fire induced high impedance faults from compromising a power supply to safe shutdown loads.
The licensee in comment 30 and Section 5.6 of Volume I of the Appendix R Engineering Evaluations stated that no loads are to be stripped for consideration of high impedance faults because of adequate calculated margins.
The staff agrees with this approach.
There is one inspector follow up item and one unresolved item identified in BFN IR 89-13 in the area of cewen bus concerns.
The inspector follow-up item deals, with power circuit coordination and the unresolved item deals with fusing of control circuits.
These items will be closed before the restart of Unit 2.
5.2
~RH S1 Spurious signals are addressed in Section 2.4.14.5 of the SE; TVA comaents 31, 32, 33, 34 and 35; and Sections 5.3.1, 5.3.2, 5.3.3, 5.3.4, 6.1.6.1, and 8.2.4 of Volume I of the Appendix R Engineering Evaluations.
12 As a result of the TVA co@vents and the Appendix R Engineering Evaluations, the first two paragraphs in Section 2.4.14.5 of the SE should be revised as follows:
"Fire in some areas can impair safe shutdown due to fire-induced spurious operation of safe shutdown equipment or components.
Therefore, Appendix R Item III.G.2 separation requirements have been provided between redundant safe shutdown equipment, components and associated cabling as far as practicable (e.g not provided in the control building) to ensure the availability of minimum SSDS.
When this is not possible.
spurious operations are to be mitigated in a timely manner by corrective manual operations at local stations.
This is performed by isolating the minimum SSDS needed from the fire affected area and providing manual control capability at the local stations.
The "licensee has identified all potential spurious oper-ations, their effects, and mitigation procedures if needed.
These spurious operations involve the RHR suppression pool suction valves, RHR inboard and outboard injection valves, RHR heat exchanger outlet
- valves, and single normally closed valves on the branch connections to the main condenser and the radwaste system downstream of the cordon valve.
The licensee has also provided analyses to demonstrate that (I) spur-ious operation of one iMSRV will not compromise the safe shutdown capability, and (2) spurious operations of high pressure make-up
- systems, such as the control rod drive (CRD), and reactor core iso-lation cooling (RCIC) systems are bounded by the spurious operation of the HPGI system.
The 1'icensee states that the spurious operation of the HPCI system, if it occurs, will be manually corrected in a timely manner by closing the HPCI steam supply shutoff valve within 10 minutes after manual scram."
This revisiun serves the purpose of adding RHR inboard injection valves to the list of spurious operations concerns and removing the operation of the feed water system as an event which is bounded by spurious operation of HPGI.
Spu-rious operation of the feedwater pumps for an interval long enough to inject water into the main steam lines is not credible as explained in Section 6.1.6.1 of Volume I of the Appendix R Engineering Evaluations.
In addition, the licensee proposes to remove power for one of the valves (FCV-74-47) in the suction line of the RHR in the shutdown cooling mode.
Also the pipe size limit for high/low pressure interface inventory loss considera-tions is one inch or less.
All of the proposed actions and modifications are considered to be acceptable by the staff.
In addition, these changes to the staff SE do no affect the staff's conclusion that the licensee's proposed handling of spurious signals is acceptable.
13
- 5. 3 Common Encl osures This was not addressed in the staff SE.
After examining switch gear equipment, distribution panels, and cabinets that provide a cozen enclosure for safe shutdown cables along with non-safe shut-down cables, the licensee determined that in general, the safe shutdown ci r-cuits were protected by circuit breaker s, fuses or isolation devices.
A major exception was the 250 volt DC breaker control circuits for the 4KV and 480V shutdown boards.
These circuits use 35 and 30 amp fuses which are considered too large for protection of No.
12 cables.
This condition was corrected by using the proper fuses.
6.0 APPENDIX R RELATED TECHNICAL SPECIFICATIONS By letter of April 14, 1989 (C.H.
Fox to U.S.
NRC),
TVA submitted Technical Specification (TS) No. 268 - Appendix R Safe Shutdown.
The purpose of this submittal is to add License Condition 2.C.5(a) and appropriate administrative controls needed to implement the BFN Unit 2 Appendix R Safe Shutdown Program.
. The BFN Unit 2 Appendix R Safe Shutdown Program was transmitted to NRC under a
separate cover letter on the same date.
The staff's review of the proposed TS is being covered as a separate licensing action (TAC No. 72965).
The licensee has been advised that the proposed TS submittal will require a revision to address the staff's concerns on strength-ening fire watch requirements in certain fire areas that do not have automatic fire protection features (other than Fire Area 9'nd the cable spreading room).
The staff's review of TS 268 covered under TAC 72965 continues to remain open until the licensee's revised submittal is reviewed and approved.
7.0 NFPA CODE DEVIATIONS This was not addressed in the staff SE.
The licensee submitted'a summary of deviations from National Fire Protection Association (NFPA) codes at BFN on August 3, 1988 (letter from R. Gridley to USNRC).
TVA concentrated their review in four specific areas:
a.
Sprinkler, Mater Spray, and Foam-Water Sprinkler Systems (NFPA 13, NFPA 15, and NFPA 16A).
b.
Carbon Dioxide Systems (NFPA 12).
c.
Fire Protection Water Supply System (NFPA 20, NFPA 24, and NFPA 14).
d.
. Fire Detection Systems (NFPA 72D and NFPA 72B).
14 Most of the deviations as described in the licensee's letter dated August 3, 1988 or through telephone conversations with the staff were accepted to the staff.
The deviations originally found unacceptable by the staff were:
a.
Lack of air supervision on the preaction system in the Reactor Building.
b.
Lack of air supervision on the preaction system in the Intake Pumping Station.
c.
Various deviations on the preaction system in the cable spreading room.
By letter of February 3, 1989 (R. Gridley to USNRC), TVA coomitted to install air supervision on the Unit 2 Reactor Building preactfon sprinklers.
The staff visited the site on March 21, 1989 (See NRC letter dated June 6, 1989) for the purpose of reviewing the deviations in the intake pumping station and the cable spreading room.
The staff determined that a1r supervision in the 1ntake pumping station could be deferred until after restart.
The staff also determined that the existing preaction sprinkler system in the cable spread1ng room was acceptable for Unit 2 restart.
TVA has,
- however, decided to replace the existing ceiling level sprinkler heads with quick response spr1nkler heads prior to Unit 2 restart (M. J.
Ray to USNRC, dated August 7, 1989).
In the February 3, l989 letter, TVA also comnitted to'performing tests and revising surveillance instruct1ons before the staff's Appendix R aud1t.
All of these items associated with the NFPA code deviations were inspected and found to be acceptable (BFN IR 89-28).
This closes the staff's action covered under TAC 00459.
8.0 AREAS OF CONCERN In the conclusion of the staff SE, the staff identified certain areas of concern which prevented the staff from concluding that the safe shutdown capab1lity (existing and proposed modifications) at the BFN units satisfies the requirements of Section III.G.2, III.G.3 and III.Lof Appendix R.
The areas of staff concern, as stated in the staff SE, are the following two paragraphs:
Protection of one train of instrumentation for suppress1on pool level and
~
temperature for the long term is an area of concern.
The proposed portable instrumentation fs only acceptable as as interim measure, provided the reliability and dependability of the instrumentation is verified at the time of the Appendix R Site Audit (SE Section 2.4.9).
At the time of the Appendix R Site Audit. the staff will verify the following:
(1) acceptability of the design basis requirements, (2) adequacy of some fire protection systems, (3) adequacy of the procedures for the testing of fire detection and suppression
- systems, (4) validity of l1censee's designation of some fire locations as solely III.G.1 areas per Appendix R cr1terion, (5) feasibility of performing with1n a very short time (e.g.,
10 minutes after manual scram) a few local manual actions (SE Section 2.4.3),
and (6) review of
15 the licensee's clarification of entry conditions for shutdown procedures (SE Section 2.4.15).
These concerns have been addressed for Unit 2 operation in this supplemental SE.
The use of portable instrumentation is accepted by the staff in Sec-tion 2.5 above.
The issues identified for the Appendix R site audit were reviewed in NRC Inspections 89-13 and 89-28 and the adverse inspection findings needed to be resolved before the restart of Unit 2 are discussed in Sections 2.1, 2.6, 2.7, 3.0 and 5.1 of this supplemental SE.
This supplemental SE only discusses and evaluates fire protection for Unit 2 operation including certain ca@non equipment in Units 1 and 3.
Is is assumed that Units 1 and 3 are shutdown with only unirradiated fuel fn their cores.
The staff concludes, based on the SE and this supplement.. that with the resolu-tion of the restart inspection follow-up items in Section 2. 1, 2.6, 2.7, 3.0 and 5.1 given above the safe shutdown capability (existing and proposed modifi-cations) at Browns Ferry for Unit 2 operation satisfies the requirements of Section III.G.2, III.G.3 and III.Lof Appendix R requirements.
The exemptions requested from certain provisions of Section,III.G.2 and III.G3 were issued by staff letter dated October 21, 1988.
- 9. 0 CONCLUS ION This safety evaluation is a supplement to the staff SE issued on December 8, 1988.
As a supplement, it addresses new information since this-SE was issued which is provided in the following documents:
(1) Volume I of the Appendix R
Engineering Evaluations for Browns Ferry in the TVA's letter dated February 14,
- 1989, (2) comments on the staff SE from TVA in its letter dated March 10, 1989, and (3) the two staff's Inspection Reports 50-259, 260, 296/89-13 and 89-28.
This safety evaluation addresses the staff's fire protection analysis for only the operation of Unit 2.
10.
REFERENCES a.
December 8, 1988. Letter from S. Black to 0. Kingsley, Browns Ferry Nuclear Plant Units 1, 2, and 3 - Appendix R Safe Shutdown System Analysis.
b.
October 21, 1988, Letter from S. Black to S.A. Mhite, Browns Ferry Nuclear Plant Units 1, 2, and 3 Appendix R Exemptions.
C ~
March 10, 1989, Letter from R. Gridley to USNRC Browns Ferry Nuclear Plant (BFN) - Appendix R Audit Ready Date, Procedures, and Safety Evaluation (SE)
Co@vents.
d.
February 14, 1989, Letter from R. Gridley to USNRC, Browns Ferry Nuclear Plant (BFN) - Unit 2 Appendix R.Audit (transmitted SSI's Appendix R
Engineering Evaluations, and Drawings).
16 e.
9>>
k.
m.
June 14, 1989, Letter from M.J.
Ray to USNRC, Browns Ferry Nuclear Plant (BFN) - Appendix R Documents Requested by NRC (transmitted revised Appen-dix R Engineering Evaluations and related documents).
August 1, 1989. Letter from B.
D. Liaw to 0. Kingsley, BFN Inspection
- Reports, 50-259/89-13, 50-260/89-13 and 50-296/89-13.
September 15, 1989, Letter from B.
D. Liaw to 0. Kingsley, BFN Inspection Report 50>>259/89-28, 50-260/89-28 and 50-296/89-28.
April 13, 1975 ~ "Plan for Evaluation, Repair, and Return to Service of Browns Ferry Units 1 and 2 (March 22, 1975 Fire)".
November 21, 1986, Letter from Gridley to Muller Browns Ferry Nuclear Plant (BFN) - 10 CFR 50 Appendix R (transmitted revised exemption request).
April 14, 1989, Letter from C.
Fox to USNRC, Browns Ferry Nuclear Plant (BFN) Unit 2 - Appendix R Safe Shutdown Program.
April 14, 1989, Letter from C.
Fox to USNRC Browns Ferry Nuclear Plant (BFN) - TVA BFN Unit 2 Technical Specification No 268 - Appendix R Safe Shutdown.
August 3, 1988, Letter from R. Gridley to USNRC, - Summary of Deviations from National Fire Protection Association (NFPA) code.
December 14, 1988, Letter from S. Black to 0. Kingsley - Browns Ferry Nuclear Plant Deviations from the National Fire Protection Association (NFPA) codes.
n.
February 3, 1988, Letter from R. Gridley to
- USNRC, Browns Ferry Nuclear Plant (BFN) - Deviations from the National Fire Protection Association NFPA) codes.
0>>
p>>
August 7, 1989, Letter from M. J.
Ray to USNRC, NFPA Code Deviations in Relation to Cable Spreading Room Sprinkler s.
October 31, 1989, Letter from M. J.
Ray to USNRC, Comnitment to Replace Fire Wall.
Principal Contributor:
Rex Wescott Dated:
November 3, 1989