IR 05000259/1989013
| ML18033A906 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 08/01/1989 |
| From: | Hubbard G, Watt J, Wescott R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML18033A905 | List: |
| References | |
| 50-259-89-13, 50-260-89-13, 50-296-89-13, GL-86-10, IEIN-84-09, IEIN-84-9, NUDOCS 8909010140 | |
| Download: ML18033A906 (30) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION TYA PROJECTS DIVISION Report Nos:
50-259/89-13, 50-260/89-13 and 50-296/89-13 Docket Nos:
50-259, 50-260 and 50-296 License Nos:
DPR-33, DPR-52 and DPR-68 Licensee:
Tennessee Valley Authority 6N 38A Lookout Place 1101 Market Street Chattanooga, Tennessee 37402-2801 t
Facility Name:
Browns Ferry Nuclear Plant Units 1, 2 and
Inspection Conducted:
May l-5, 989 Inspt.ctcrs:
.ex G. 'nescott, earn LI'a,:er
~aj l~s'
%PC I
James
.
+at Date
Consultants:
Approved by:
E. NacDougall and R. Hodor, Brookhaven National Laboratory
/ j eor
.
ub ar,
se 5t Plant Systems Branch TVA Projects Division Office of Nuclear Reactor Regulation 890V0>0>no 8e080>
PDR ADOCK 05000259
= PIIIU=.
SUMMARY
~Sco e:
This special announced inspection was conducted in the area of postfire safe shutdown to evaluate Unit 2 for compliance with Section III.G and III.L of Appendix R to
CFR 50.
Compliance with III.J (emergency lighting) and fire protection related aspects of III.G will be evaluated in a later inspection.
Also this inspection is specifically for Unit 2 restart, with Units 1 and
shutdown and not having returned to criticality from the present outage.
The basis for this inspection is the staff SE on Appendix R safe shutdown systems analysis issued on Oecember 8,
1988, the staff exemption SE issued on October 21, 1988, TVA's comments on the staff SE's dated March 10, 1989 and Volume I of TVA's Appendix R Engineering Evaluations.
Resu1ts:
One unresolved item was identified:
50-259/89-13-05, 50-260/89-13-05 and 50-296/89-13-05:
Potential failure to meet Section III.G.3 of 10 CFR 50, Appendix R, Paragraph 3.1.2.
Nine inspector followup items were identified:
50-259/89-13-01, 50-260/89-13-01 and 50-296/89-13-01:
Correction to sheet
of Appendix R cable listing with respect to the status of valve
PCV 69-14, paragraph 2.1.
50-259/89-13-CZ, 50-260/89-13-02 and 5G-296/89-13-02:
Cor rection of drawing error for corduit ZIE 174 ir Fire Area
'.8, paragraph 2.1.
'0-259/89-13-03, 50-260/89-.'>-03 and 50-'96g89-13-0.":
l'ocumentat on of,'l'aC requirements in postfire instructions, pa. agraph 2.6.
50-259/89-13-04, 50-260/89-13-04 and 50-260/S&-13-04:
Listing of ma
forum fault currents on all breaker fuse coordination sheets, paragraph 3.1.1.
50-259/89-13-06, 50-260/89-13-06 and 50-296/89-13-06:
Completion of work packages for 5 circuit breaker modification packages, paragraph 3. 1.3.
50-259/89-13-07, 50-260/89-13-07 and 50-296/89-13-07:
Placement of additional emergency light to illuminate the remote shutdown panel, paragraph 4. 1.
50-259/89-13-08, 50-260/89-13-08 and 50-296/89-13-08:
Modification of SSI's to clearly indicate a change of location for operations, paragraph 4. 1.
50-259/89-13-09, 50-260/89-13-09 and 50-296/89-13-09:
Proper tagging of valves needed for manual actions, paragraph 4. 1.
50-259/89-13-10, 50-260/89-13-10 and 50-296/89-13-10:
Oocumentation of modifications identified by TVA to the inspection team relative to changes to the SSI's and Appendix R Engineering Evaluations, paragraph f
1.0 PERSONS CONTACTED Licensee Em lo ees
- James East, Licensing Engineer
- Patrfck Carfer, Site Licensing Manager
- James E. Swindell, Plant Support Superintendent
- John Hutton, Operations Superintendent R.R. Baron, gualfty Assurance Engineer
- Robert T. Wimbrow, Nuclear Engineer Robert M. Tuttle, Site Security William R. Lagergren, Jr., Protective Services
- J.J. Pierce, Engineering Specialist Appendix R
- Harvin Belew, Engineering Specialist - ICC
- Tom Temple, Engineering Specialist - Systems
- B. F. Miller, Operations
- Charles Anderson, Special Projects Engineer R. Burt, Construction
<<Jim Kern, Ffre Protection Carry Parvin, guality Engineering Hike Coston, guality Engineering Cecil Thomas, Electrical Engineer
- C.E. Massey, Appendix R Project Manager J.
McCarthy, Regulatory Licensing Supervisor
- Tom Davis, Site Fire Protection/Appendix R
- Robert Williams, Senior Electrical Engineer
- Julio Gomez, Electrical Engineer
- Raymond Wright, Principal Electrical Engineer
- R. E.
Young, Modifications
- T.R. Bradish, PORS Superfntender t P.
Ebersole, Mechanical Engineer Other Or anizations
- W. Shields, Bishop, Cook, Purcel'rd Reynolds
- J.
Gadow, NEC J. Nicely, Contractor
- James Straussen, Toledo Edison Co.
- Mike Lalon, Toledo Edison Co.
- Stuart Kamer, Toledo Edison Co.
NRC Personnel
- R. Pierson, AD/TVA Technical Programs
- D. Carpenter, Senior Resident Inspector
- E. Christnot, Resident Inspector
- Attended exit interview
2.0 SAFE SHUTDOWN PERFORMANCE GOALS AND SYSTEMS This part of the inspection consisted of an assessment of the licensees compli-ance with the Appendix R requirements for protection of redundant systems, alternative shutdown capability, operator training and operator actions.
On a
sample basis, the inspectors examined measures that the licensee implemented to assure safe shutdown capability and compliance with Appendix R.
2. 1 S stems Re uired for Safe Shutdown The systems required for Appendix R safe shutdown for Browns Fer ry Nuclear Plant (BFN) Unit 2 include:
Residual Heat Removal (RHR) system Main Steam Relief Valves (MSRV)
Residual Heat Removal Service Mater (RHRSM)
High Pressure Coolant Injection (HPCI)
The HPCI system is used for gradual depressurization and reactor coolant makeup for a fire in any fire area except Fire Area 2 (Unit 2 Reactor Building), 9 (4 KV shutdown Board room C),
16 (Control Building elevation 593, 606 and 617)
and 18 (Unit 2 Battery and Battery Board rooms).
The HPCI system can maintain the unit in Hot Shutdown for up to two hours.
After this period the reactor is further depressurized with the NSRV's to cold shutdown which is maintained with RHR in the low pressure coolant injection (LPCI) mode.
For a fire in Fire Areas 2, 9, 16 and 18 the HSRV's are used to rapidly depressurize the reactor and the RHR system in the LPCI mode is used to maintain cold shutdown.
t The RHRSW system is used for heat removal from the suppression pool.
During the inspection fifteen safe shutdown components were selected out of the list of components required for fire in fire areas 2-4 (zone 4 of Unit 2 Reactor Building),
16 (Control Builaing),
18 (Unit 2 Battery Board Room)
and 25 {Turbine Buildir:g and Intake Structure).
The routing cf about 70 pnwer and control circuits necessary for these components were verified ~rom draw>ngs to assure that none of the required circuits passed thrcugh the fi"e area (without protection) for which the component was in ;he shutdown path.
A sample of the cables that were verified is listed below,:
~Com anent RHR SM A3 RHR SM B3 2-FCV-074-067 Cable Number 3E51580-I E52575-II 2E51050- I 2E53579-II 2E53583-II 2E53906-II 2E53907-II 2E53908-II 2E5769-I
2LI-003-058A 2PCV-001-004 2E51555-I 2E532-I 2E533-I 2E534-I 2E448-I 2E559-I 2E5174-I 250V DC RMOVBD-2B RMOVBD-2C 480V SWBO-1A 480V SWBD-2A 480V SWBD-3A 2B80IES 2B84I ES 1B83>>IE 2B284-IE 3B88-IE All cables and circuits inspected were found to be acceptable.
However, the Appendix R cable list needs updating on sheet 54 in regard to the status of valve 2PVC 69-14.
This valve was listed as required for Fire Area 16 (Control Building) and required cables being routed there.
This valve is not listed as required for shutdown for any area in the Appendix R Mechanical Equipment Calculation.
This issue is identified as Inspector Followup Item', IFI 50-259/89<<13-01, 50-260/89>>13-01, 50-296/89-13-01.
In addition, selected conduits and cables were field verified for proper routing.
One of the selected cables, 2IE 174, was incorrectly shown as not passing through Fire Area 18 (Unit 2 Battery Branch Room)
by one of two drawings showing the routing of the cable.
The other drawing correctly showed the cable routing.
This issue is identified as Inspector Followup Item IFI 50-259/89-13-02, 50-260/89-13-02, 50-296/89-13-02.
This item requires correction and analysis of root cause.
2.2 Control of Reactivit Upon de:ection of a disabling 'ire requirirr shutdown, ch< contru'cns will be
- nserted using the scram buttons in the cc,nvrol room or automatically by l.he reactor protection system (RPS)
upon loss of of--site power.
This is true for a fire in any fire zone or area.
The inspectors verified tha-scram can also be accomplished outside of the control room by opening the RPS motor generator (MG) set output breakers.
There were no adverse inspection findings in this area.
2.3 Reactor Coolant Pressure and Inventor Control For those fire areas which do not use the MSRV/LPCI method of shutdown, the HPCI system is used for reactor coolant make-up.
The source of make up water will be the condensate storage tank (CST) except for a fire in Fire Area 25 in which case suction is taken from the suppression pool.
A fire in Area 20 which contains cabling for CST suction valve PCV-40 could result in valve closure and loss of suctio The worst case alternative shutdown scenario involves the availability of the minimum safe shutdown systems which include only the manual control of the MSRV's, RHR in the LPCI mode and the RHRSW systems.
The MSRV's are manually operated to reduce the reactor pressure to allow low pressure injection (below 450 psi).
During the depressurization there is a short term uncovery of the core with a resultant increase in the temperature of the top fuel node prior to reflooding with LPCI.
The Licensee requested and was granted an exemption for temporary uncovery of the core (October 21, 1988).
Pressure control for both shutdown methodologies relies on the MSRV's initially to prevent vessel overpressure and for depressurization which is accomplished by manual actions.
A minimum of three MSRVs is required if the HPCI mode is not entered.
In order to insure a long term MSRV air supply beyond the two and one-half hours on the reserve tanks of the drywell air control system, the licensee has provided a backup from the containment atmosphere dilution (CAD) system.
This consists of a new line with check and isolation valves installed from the CAD nitrogen supply trains to the drywell control air header.
The backup supply from the CAD is connected to allow the accumulators to be recharged by manual action at panel 25-246 B after two hours into the Appendix R scenario.
The design of this system was reviewed and found to be acceptable (Drawing 47W230-13, Revision 0).
2.4 Removal of Deca Heat In addition to providing low pressure makeup, the LPCI mode of RHR removes decay,'heat from the suppression pool by pumping suppression pool water through the RHR heat exchangers and transferring decay heat to the RHR Service Water.
The cooled water is injected into the pressure vessel which eventually floods to the level of the steam lines and MSRV's, scme of which are opened to provide a return pa~.h
-.o tne suppression pool, The structural integri:y of the steam lines and their supports during :h~s alternate shutdowr, cooling mode was demonst>ated by ana!yses and tests.
This was documented by a September I, 1983 letter
=rom l.'yle Labs which was audi:ed by the inspeczion team.
There were no adverse irspection findings in this area.
2.5 Process Monitorin The required process monitoring instrumentation for the reactor consists of reactor pressure, reactor level, suppression pool temperature, suppression pool level, dry well temperature and dry well pressure.
The inspectors verified that the process monitoring instrumentation available in the main control room and at the remote shutdown panel meets the requirements of Generic Letter 86-10 and IE Information Notice 84-09.
The inspectors also reviewed the surveillance requirements for Appendix R related controls and instrumentation on the remote shutdown panel.
These were found to be acceptable.
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2.6 Su ort S stems and E uf ment For the safe shutdown systems described in the pr eceeding Sections of 2.0, the following support systems are required to be available:
Auxiliary Power Supply System (APS)
250 Volt D.C.
Power System Emergency Equipment Cooling Water System (EECW)
Diesel Auxiliary Systems MSRV Control Air Supply HVAC System For some fire areas at BFN, only one of the two EECW pumps required for diesel generator cooling fs assured of auto starting.
The other EECW fs required to be manually started within ten minutes of diesel generator start.
The inspection team requested BFN to provide supporting documentation showing that under those conditions, the jacket cooling water temperature would not be exceeded.
Calculations and test results that demonstrated that the jacket water design limit of 208'F would be reached if no cooling water was provided within 16 minutes (unloaded)
and 5 minutes (loaded)
were provided to the team and reviewed.
The one EECW pump is available upon diesel start and furnishes approximately 80K of the cooling flow for the diesels until the second pump fs started within the 10 minute time frame.
This item was found to be acceptable by the inspectors.
BFN personnel also furnished an analysis (B 22 870717 004) in response to the inspectors concern that operation of the one auto started EECW pump could induce pump runout conditions prior to starting the second pump manually.
The analysis was reviewed and found to be acceptable by the inspection team..
The availability of the RHR pumps at BFN has been affected by the fouling of the shel';
side of the seal cooling heat excnai.gers.
The fou~ing of the i!eat exchangers has resulted
!n frequent mafntenanrd with the pumps out of'erv ce.
Tile licensee has propnsea operatirg the RHR purp without EECW flow if ~o! 1!ng cccvrs during cn Appendix R event and providea
- he inspectors with a copy of, calculation MO-f0074-87017 July lG, 1987 for review.
The shutdown ceo,irg mode of RHR pump operation wi thout EECW cooling (tube side seal water flow must be maintained)
can result in an initial water/sea'.
.emperature of 300'F which is gradually reduced to '25'F in approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.
An evaluation by the General Electric Plant Materials and Technology Group states that it would be permissible to run the RHR pump without seal coolers at temperatures up to 215'F indefinitely with no adverse effect and at temperatures between 215'F and 400'F for whatever time fs required during an emergency.
If the pumps are operated in the higher range (over 215') the seals should be inspected for wear.
The inspectors found this mode of RHR operation acceptable during an emergency.
An item not yet completed by the licensee is the inclusion of HVAC requirements into the post shutdown procedures.
This is identified as Inspector Followup Item IFI 50-259/89-13-03, 50-260/89-13-03, 50-296/89-13-03.
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2.7 Manual Actions Alternative Shutdown At BFN Unit 2, alternative shutdown capability is provided for areas of the plant that do not meet the requirements of III.G.2.
These areas include the cable spreading room and the control room which are included in Fire Area 16.
The Unit 2 remote shutdown panel 25-32 serves as a
command center following a control room evacuation for remote shutdown.
The panel provides instrumentation to monitor reactor pressure and level and suppression pool level and temperature.
Local control of four MSRV's used for rapid depressurization is provided at the panel.
Other operators perform manual actions at the 4160 volt shutdown board, 250 volt Reactor Motor Operated Valve (RMOY) Board, 480 volt shutdown board and locally operate valves.
Procedures and training are discussed in Sections 4. 1 and 4.2.
The adequacy of the remote shutdown instrumentation and controls was reviewed by the inspectors and is addressed in Section 2.2, 2.3 and 2.5.
2.8 Modifications The inspection team selected the following samples of modifications based on their importance to safe shutdown.
An electrical review was made of these modifications.
Modifications-Electrical Modification
~Setem Fire Area Comments Circuit Modifi-cation 8.2.2 2 Circuit Modifi-cation 8.2.3 RHR 16-2/4
Review of modifi~ation unsatatisfactory Unsatisfactory.
FCV-52 Circuit ModiA-cation 8.2.5 RHR Several Satisfactory - Intertie DCh~4675A Cable Modifi-cation 8.3.2(b)
Cable Modifi-cation 8.3.2(c)
Cable Modifi-cation 8.3.2(d)
Cable Modifi-cation 8.3. 1 RHR RHR RHR MSRVS
2/4
18 Satisfactory Satisfactory Satisfactory Satisfactory - new conduit route not complete The concern identified during the modification review pertained to Modification 8.2.2 and 8.2.3.
A lack of fuse coordination was demonstrated during testing when the primary control transformer fuse blew instead of the secondary control transformer fuse.
This concern is addressed in more detail in Section 3. 1.2.
Otherwise, the inspection team found the modifications were well tracked and corrective action was put in effect when needed.
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See Section 3.1.2-6-
3.0 ASSOCIATED CIRCUITS 3.1 Common Bus Concern The cornrron bus concern was approached from the context of coordination of power circuits, coordination of control fuses, and the high impedance fault analysis.
The philosophy of the BFN plant was to review all Class IE circuits and all associated circuits and all non-Class 1E circuits that are routed through a
Category I structure.
3. 1. 1 Power Circuit-Coordination Both Bechtel and TVA have performed power circuit coordination studies.
The inspectors reviewed samples of both during the inspection.
From the Bechtel Coordination Analysis ED-f2999-880562, Revision 1, several 4kV and 480 volt shutdown board circuits were reviewed and found satisfactory.
The cables selected for the staff audit were K322 and K321, which run through fire area
from the diesel auxiliary board A and B, respectively.
The licensee demonstrated the coor dination of these cables by calculation No.
ED-(2999-880562, dated November 2, 1988.
Circuits involving cables 2B84, 2B80 and 2B95 were addressed in TVA calculation ED-(2000-87-550.
At first the main fuse upstream of the main circuit breaker appeared to be uncoordinated relative tothe single line diagram for curve DC-4.
As a result, the inspection team and licensee agreed that showing fault currents for both the fuse and circuit breaker on the curves could increase clarity in future analysis.
This will be done for all curves where such infor-mation was omitted in the past.
This item is identified as Inspector Followup Item IFI 50-259/89-13-04, 50-260/89-13-04, 50-296/89-13-04.
The inspection team reviewed the coordination of cables 2880 IES, 2B84 IES, the feeder-cable for SWBDA (28951E),
the feeder for RHRSW pumps C-l, the 2FCV74-53 and other'ircuits chosen at random during tte audit.
These were fourrd to be acceptable.
3. 1.2 Fusin of Control Circuits In order tc review the coordination of fuses, the staff selected the following control circuits for audit:
a)
Circuit Modification 8.2.2 for disabling control circuit of RHR valves, (2FCV74-53 and 67)
b)
Circuit Modification 8.2.3 for RHR outboard injection valve without backup control, (2FCY74-52 and 66).
c)
Coordinating fuse for 250Vdc control circuits.
d)
Fuse replacement 250Vdc control circuits.
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As a result of the review the following was identified in regard to fusing of control circuits:
some control fuses did not match the sizes shown in the schematics some control fuses were not coordinated as found in modification testing for some control circuits, a fuse coordination study was not per-formed for Appendix R
The unresolved concern regarding BFNP fusing is that for Unit 2 FCV74-52, 53, 66 and 67 as shown on schematic 2-45E779-22, REV3, a coordination study was not performed for Appendix R review.
A coordination study completed during the audit showed that the circuits were satisfactorily coordinated.
The control transformer primary fuse was incorrectly shown as a six amp fuse when it was actually a three amp fuse.
However, the updating of the prints is in process and under the control of the above noted procedure.
The staff position is that the root cause for the lack of a coordination study should be determined and action taken on that root cause.
This concern is identified as Unresolved Item URI 50-259/89-13-05, 50-260/89-13-05, 50<<296/89-13-05.
3.1.3 Hi h Im cdance Fault Anal sis The methodology of the high impedance fault analysis found in ED f000-87077 was examined and found to be conservative and satisfactory.
However, five circuit breaker settings on Page 109 of ED-(}000-87077 dated January 31, 1989 have not yet been implemented.
These corrections are as follows:
increase feeder breaker margins for 480V CB Vent Bd B
480V CB Vent Bd B
4CGV CB Vent Bd B
480V Corr;.n Bd
Compt iD Co(i.'pt 70 Compt 3C Change L
set to.9x (was
.Gx)
Change LT se to.gx (was
.Bx)
Change current senscr to 500A, set LT vo
.Gx Increase feeder breaker margins for 48OV RMOV Bd 2A 480V RMOV Bd 2A 480V RMOV Bd 2A Compt 3D Compt 15D Change LT set to 1.3x (was 1.2x)
Change LT set to 1.2x (was 1.2x)
There did not appear to be any mechanism in place to assure that these corrective actions would be implemented.
The staff position is that the root cause for these omissions should be determined and corrective action taken.
This item is identified as Inspection Followup Item IFI 50-259/89-13-06, 50-260/89-13-Q6, 5Q-296/89-13-06.
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The spurious signals caused by a fire are also addressed in Section 3.5 (isolation devices)
and Section 2.8 (modifications) of this inspection report.
In order to isolate systems and components for fire instigated spurious signals at BFNP, manual actions are taken.
These manual actions have been addressed in Section 2.7.
The spurious actions taken to isolate Hi-Lo pressure interface components are summarized in Tables 5-3 and analyzed in Table 5-2 of the licensee's Appendix R engineering evaluations.
In most cases no action needs to be taken due to fire effects.
Motive power is removed for RHR 2FCV-47 and 48.
The power cable for 2FCV-47 was rerouted in conduit, and its breaker tagged out during normal operation.
The above actions were reviewed by the inspectors and found to be acceptable.
3.3 Cowmen Enclosure The inspectors reviewed common enclosure associated circuits to determine if redundant circuits are routed together in a raceway or enclosure and are not electrically protected, or if a fire can destroy both circuits due to inadequate fire barriers.
The same cables associated with the Power Circuit coordination review (K321 and K322) were selected for the common enclosure review.
Both cables run through fire area 18 and supply loads required for safe shutdown.
Cable 2BBOIES which runs from battery board 3 to 250V DC MOV BD 2B was also selected.
According to the licensee, there are about 70 cables at BFN that are non-class lE cables that do indeed run in both Divisions 1 and 2.
However these have been analyzed and double fuses have been or will be, put on each of the circuits.
A random review of ten of these circuits was performed to verify the double fuse installation.
All were satisfactory.
An FSAR Amendment is being processed to document the licensee's position and actions for common enclosure circuits.
3.4 Rela and Circuit Breaker Calibration The inspection tea, interviewed the eiectrical technicians and electrical engineers in charge of the maintenance of relays and circuit breakers.
The power relays at BFN are calibrated at a frequency of once every four years.
The 4kV shutdown/load shedding relays are calibrated every six months.
Al)
BFN power relays have a status report every quarter.
Relay procedures DPSO-SMI-I-B-4, Revision 4 and DPSO-SMI-3-B-4, Revision 2 were reviewed for staff and QA sign off.
The power and molded case circuit breakers are tested at each refueling outage at a maximum.
Other circuit breakers are tested annually or bi-annually depending on their function and availability.
The inspection team considered these frequencies to be acceptable.
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3. 5 Isolation Device The inspection team reviewed Appendix R related isolation devices at BFN and found them to be acceptable for the following reasons:
All Class 1E isolation devices used at Class 1E interfaces were them-selves Class lE.
For fire area 16 existing isolation devices were used; For other fire area's new Class 1E-GE control switches (SBM)
matched those previously in use at the plant.
Key locking was used at the transfer point for RHR/LPCI valves 1FCV's 74-53 and 67.
All other isolation transfer points were protected by annunciator alarm in the control room.
All isolation devices used for Appendix R requirements were manually operated.
4. 0 PROCEDURES TRAINING 4.1 Procedures BFN has developed safe shutdown procedures for Unit 2 safe shutdown in the event of a disabling fire in any of the 25 designated fire areas.
Procedure No. 2-SSI-001 (safe shutdown instruction) is intended for safe shutdown for fires in those fire areas where the HSRV/LPCI methodology of shutdown has been designated.
It is a general procedure that defines the entry conditions for those fire areas using the MSRV/LPCI methodology (Fire Areas 2, 9, 16, 18).
Once entry conditions are met, separate subinstructions are provided for each of the preceding fire areas (2SSI-2, 2SSI-9, 2SSI-16, 2SSI-18).
Procedure 2-SSI-i6 contains the steps necessary to shutdown Unit 2 in the event of a disabling fire in the control building and the use of alternative shut-
'own.
Attachments A, B, C, 0, E ar!d F to the procedure contain the instructions foi the manual actions renuirea for sa>e shutdown, i~t the initiation of the shutdown, five operators are assigned by the shift superv',sor to perform the manual actions <<t different locations with each oper tor taking an attachment (A, B, C, D, E).
Attachment F is assigned to one of the five operators after they have completed the actions in their attachment ana is not required to be completed until nine hours into the shutdown.
The shift supervisor is respon-sible for direction of the shutdown and is stationed at the backup control panel 2-25-32 with Operator No.
1.
Time lines are included for completion of individual sections of each attachment.
Other procedures that use the MSRV/LPCI methodology for safe shutdown of Unit 2 that were reviewed during the inspection were Procedure No. 2SSI-2-4 (Unit 2 Reactor Building 1one 4) and 2SSI-18 (Unit 2 Battery 5 Board Room).
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~ l Procedure 2-SSI-002 is implemented for post-fire safe shutdown for fires in those fire areas where the HPCI, MSRV, LPCI methodology has been designated.
The 2-SSI-002 procedure follows the same format as the 2-SSI-001 procedure and is used for fires in fire areas 1,
3 through 8, 10 through 15, 17, and
through 25.
All of the preceding fire areas are in Units
& 3 with the exception of Fire Area 25 (Turbine Building Units 1, 2 and 3).
The inspection team reviewed the procedures for Fire Area 25 (Procedure No. 2-SSI-25II) with licensee operations and systems personnel.
Procedure 2-SSI-25II is used for a
fire in any part of the Turbine Building, Cable Tunnel, Intake Pumping Station and the Radwaste Building except for the basement of the Intake Pumping Station where Procedure No. 2-SSI-25I is used.
The inspection team completed the review of the procedures 2-SSI-001, 002, 2-SSI-2-4, 2-SSI-16, 2-SSI-18, and 2-SS-25II, and with the exception of a few minor coments, found them to be acceptable.
Particular attention was paid during the procedure review to the sequencing of the operator actions required, and the agreement between the timelines specified in the procedure and those called for in the licensee's Appendix R Evaluation Documentation.
Dur ing the morning of May 3, 1989 the inspection team conducted a walkdown of Alternative Shutdown Procedure 2-SSI-16.
Since the BFN procedures call for many manual actions to be accomplished within a required time frame the walkdown was setup to be timed, with each inspector assigned to an operator to keep track of the time required to complete the more imnediate actions.
Critical time lines for completion of manual actions included:
First 10 minutes
-
Establishing local control of Backup Control Panel 2-25-32 (Operator No.
1)
RPS MG Set 28 Manual Trip and align power to 4160V Shutdown Board D.
Align 4160V Shutdown Board C Div. I.
RPS MG Set 2A Im RPT Division II Control Power Trip
HPCI Isolation.
First 20 minutes
-
Align Power to 4160V Shutdown Boards A u B, 480V Shutdown Board lA and start PHR pump C.
Results of the time line walkdown demonstrated that the manual actions for the ten minute requirements were completed in six to seven minutes and the
minute time line which included start of low pressure injection with RHR Pump 2C was completed in 18 minutes.
An abbreviated walkdown of the 2-SSI-25II (Turbine Building) was conducted after the 2-SSI-16 walkdown and again the operators demonstrated the feasibility of accomplishing the required manual actions within the prescribed time line ~
~
e In addition to checking the time lines for manual actions the inspectors checked for:
Operator familiarity with the procedural steps.
Adequate emergency lighting to read gauges and controls.
Proper tagging/identification of equipment and controls.
Availability of tools, protective clothing and procedures.
Adequacy of comnunication.
Feasibility of manual actions; particularly those requiring manual operation of valves.
Items identified by the inspectors during the walkdown included:
Inoperable emergency light at the backup control panel.
The licensee agreed to correct this situation.
Also, an additional emergency light was recommended by the inspectors.
The licensee concurred with this recoaeendation.
This is identified as Inspector Followup Item IFI 50-259/89-13-07$
50-260/89-13-07$
50-296/89-13-07.
During the walkdown of 2-SSI-25II, one of the operators exhibited confusion as to his plant location and his understanding of the procedural requirements.
The licensee agreed to add clarifying statements to the procedure.
This is identified as Inspector Followup Item IFI 50-259/89-13-08, 50-260/89-13-08, 50-296/89-13-08.
Another item observed during the walkdown was that some of the valves requiring manual operation were not tagged properly.
The licensee agreed to review and correct.
This is identified as Inspector Fol l owup Itern IFI 50-259/89-13-09, 50-260/89-13-09, 50-296/89-13-09.
The shift supervisor who was directing the shutdown was unable to radio communicate with Operator No.
5 at one point.
However, another operator relayed information between the shift supervisor and Operator No. 5.
The licensee indicated that the radio system was being upgraded for improved communication and was about 15% complete.
The inspectors also recommended that a test of the upgraded system be performed.
The inspectors also recommended that phonetics and proper protocol in radio communications be used by the operators.
The licensee agreed to review this area for possible training and procedural action.
Because the radio system upgrade was already being undertaken and tracked, an inspector followup item was not identified for that issue.
Despite the items noted, the inspection team determined that the manual actions required could be completed within the required time frames, the operators were well trained, familiar with the procedures and equipment and capable of implementing safe shutdown of the unit while adhering to the functional goals of Appendix R.
Following the procedure review and walkdown, the licensee furnished the inspection team with a list of upcoming changes to Appendix R
documentation.
The list was comprised of five system oriented open items which affect the SSI's and the Volume 1 Engineering Evaluation.
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The five items are listed below.
1.
Actions will be added to the SSI's to assure isolation of the reactor water cleanup system (RMCU).
2.
Actions will be added to the SSI's to properly sequence Ul/U2 4KV Bd alignments with U3 4KV Bd alignments 3.
The action of removing control power from the 4KV SD BD's was deleted from the SSI's to reduce the number of manual actions.
Seven additional Appendix R disconnect switches are being added and 20 circuit modifications are being made to support this deletion.
4.
Actions to maintain containment atmosphere dilution (CAD) valve g-FCV-84-16 closed for initial two hours will be deleted from SSI's.
Analyses showed that adequate nftrogen supply exists for MSRY's even with spurious injections into the Unit 2 drywell.
5.
An addftional load shed fsolatfon switch for Chiller 1B is needed for a fire fn Fire Area 9.
This will require changes in the SSI's.
The five items were reviewed by the inspection team with licensee operations and systems personnel to determine fn particular, the effect of these changes on the operators ability to adhere to the critical time lines.
The results of the review indicated that these changes would not jeopardize completion of manual actions within the requisite time frame.
The required modifications of the SSI's and Yolume I of the Appendix R Engineering Evaluations is identified as Inspector Followup Item IFI 50-259/89-13-10, 50-260/89-13-10, 50-296/89-13-10.
4.2
~Trainin The inspection team reviewed the operator training program at BFN with the operators, operations staff personnel and the training officer.
Operator training at BFN includes classroom instruction, walkdown uf the procedures for safe shutdown, and a written examination over a period of five days and is repeated annually.
Training records and lesson plans were also reviewed.
The lesson plans are periodically updated and important changes are transnitted formally to the operators who must sign and return them.
The training officer was questioned about the use of the simu1ator as an aid in Appendix R operator training.
The officer stated that the current simulator was of too limited features and capabilities to provide realistic reactor and system characteristics during the process of performing the manual actions.
It was stated that negotiations were currently fn process which are expected to provide a
new simulator in about two years.
The new simulator could compliment Appendix R training.
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The inspectors noted the absence of the fire battalion and incident conmander (or reference thereto) during the timeline walkdowns even though significant interactions would occur in determining the location and severity of the fire.
Once the fire zone and fire severity are specified, the Appendix R procedures are identified.
Interaction continues at least in the control room between the Shift Operators Supervisor and the Incident Comanders.
However, the fire battalion is primarily concerned with the control of the fire and the operators perform those required manual actions to achieve safe shutdown.
During the walkdowns and interviews with the operators the inspection team was very favorably impressed with the positive attitude of the operators relative to the success of the Appendix R procedures.
The operators considered their training sessions as their opportunity to verify the procedures.
It was clear that the operators were confident of the procedures and their ability to perform them within the allotted time.
The inspection team did not identify any unacceptable items in the review of operator training at BFN.
5.0 EXIT INTERVIEW The inspection scope and findings were summarized on May 5, 1989 with those persons indicated in Section 1 above.
The licensee acknowledged the following inspection findings:
One unresolved item was identified:
50-259/89-13-05, 50-260/89-13-05, 50-296/89-13-05:
Potential failure to meet Section III.G.3 of 10 CFR50, Appendix R, paragraph 3.1.2.
Nine Inspector Followup Items were identified:
50-259/89-13-01, 50-260/89-13-01, 50-296/89-13-01:
Correction to sheet 54 of Appendix R cable listing with respect to the status of valve 2PCV 69-14, paragraph 2. 1.
50-259/89-13-02, 50-260/89-13-02, 50-296/89-13-02:
Correction of drawing err or for conduit 2IE174 in Fire Area 18, paragraph 2.1.
50-259/89-13-03, 50-260/89-13-03, 50-296/89-13-03:
Documentation of HVAC requirements in postfire instructions, paragraph 2.6.
50-259/89-13-04, 50-260/89-13-04, 50-296/89-13-04:
Listing of maximum fault currents on all breaker fuse coordination sheets, paragraph 3.1. 1.
50-259/89-13-06, 50-260/89-13-06, 50-296/89-13-06:
Completion of work packages for 5 circuit breaker modification packages, paragraph 3. 1.3.
50-259/89-13-07, 50-260/89-13-07, 50-296/89-13-07:
Placement of additional emergency light to illuminate the remotes shutdown panel, paragraph 4.1.
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50-259/89-13-08, 50-260/89-13-08 and 50-296/89-13-08:
Modification of SSI's to clearly indicate a change of location for operations, paragraph 4.1.
50-259/89-13-09, 50-260/89-13-09 and 50-296/89-13-09:
Proper tagging of valves needed for manual actions, paragraph 4.1.
50-259/89-13-10, 50-260/89-13-10 and 50-296/89-13-10:
Oocumentation of modifications identified by TVA to the inspection team relative to changes to the SSI's and Appendix R Engineering Evaluations, paragraph 4.1.
All of the above items are required to be resolved prior to restart of Unit 2.
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