ML19290E897
| ML19290E897 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 03/12/1980 |
| From: | Hancock J FLORIDA POWER CORP. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 3--3-A-3, 3-0-3-A-3, IEB-79-27, TAC-42664, NUDOCS 8003170282 | |
| Download: ML19290E897 (72) | |
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++ e 0 et A T i,; N March 12, 1980 File:
3-0-3-a-3 Mr. Harold R. Denton Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 Subj ect:
Crystal River Unit 3 Docket No. 50-302 Operating License No. DPR-72
Dear fir. Denton:
Your March 6,1980, letter to all operating B&W reactor licensees, concern-ing our recent transient at Crystal River Unit 3, requested that Florida Power Corporation provide, in accordance with 10 CFR 50.54(f), its response to Items 1 through 5 of Enclosure 3 of your letter.
In addition to Items 1 through 5, Florida Power Corporation was requested to address an additional three (3) items contained in Enclosure 3 of your letter.
In that regard, Florida Power Corporation hereby submits for your review, our responses to Items 1 through 5 and the three (3) additional items spe-cific to Crystal River Unit 3.
As requested in your letter, our response to Items 6 and 7 of Enclosure 3 will be submitted no later than fiarch 17, 1980.
If you require any further discussion concerning our response, please con-tact this office.
Very truly yours, FLORID POWER CORPORATION 1
b A. Hancock Director Nuclear Operations Simpson(3/6/80RFI)DN-94 Attachments General Office 3201 Tmnwouen snee soum. P O Bcx 14042. St Petersburg Ficnda 33733 e 813 ---866-5151 8 0 0317 00s1L9s
STATE OF FLORIDA COUNTY OF PINELLAS J. A. Hancock states that he is the Director, Nuclear Operations, of Florida Power Corporation; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of his knowledge, information, and belief.
b
[
.' A. Ha hc oc k Subscribed and sworn to before me, a Notary Public in and for the State and County above named, this 12th day of March,1980.
Notary Public Notary Public, State of Florida at Large, My Commission Expires: August 8, 1983 Simpson(Notary)DN-94
t i
FPC RESPONSE TO RE0iJEST FOR INFORMATION BY HAROLD R. DENTON 01 " ARCH 6, 19RO GENERIC REQUEST 1 Summarize power upset events on NNI/ICS that have previously occurred at your plant.
RESPONSE
Encl osed, as Attachment 1, i s our summary of historical control power fail ures at CR-3.
REQUEST 2 Specifically review the Crystal River Event, and address your suscepti-bility to it in general.
RESPONSE
This request does not apply to CR-3, e-the event occurred at our pl a nt.
The infomation FPC is providing in thi s responsc, i.e.,
sequence of events, proposed corrective actions, etc., clearly identifies the impact of this event on Crystal River Unit 3, and provides a summary of the ac-tion that we propose to resolve the creblem at our plar.t.
RE')UEST 3 Set forth the infonration presented by each licensee in the meeti ng on itarch 4, 1930.
RESPONSE
Enclosed you will find the following docurents, which vere presented in our March 4,1980, meeti ng:
- 1. - Sequence of Events for the February 26, 1930, CR-3 Transient, Rev. 5.
- 2. - Copies of overhead slides used with the Sequence of Events.
- 3. - Description of Equipnent Deficiencies.
Simpson(3/6/80RFI)DN-94
Response 3 (Continued)
- 4. - Crystal River Unit 3 Feedwater Oscillations.
- 5. - Corrective Action List (The Corrective Action Lists for prior-to-startup, and for the next ref ueling outage, that were presented at the March 4,1980, meeting, have been combined a rd expanded to include the B&W r ec ommenda tions. )
The Long-term Corrective Action List has al so been expanded since our March 4,1980, presentation.
REQUEST 4 Address infonnation available to the operator following various fiNI/ICS power upset events, including:
how the operator knows what's good information what information is needed for cold shutdown.
RESP 3!1SE N'il Power Loss X or Y 1.
Fail ure Mode of indicators arti recorders.
a.
Loss of DC Bus:
(1)
Iridicators and recorders fail to mid scale for the most part.
Recorders continue to turn (chart).
b.
Loss of inco,i ng AC (1)
Indicators will fail to nid scale.
(2) Recorders will fail in same posi tion as before powe r loss, i.e., i f recorder i.e at 70% before loss of AC, it will remain at 70L Chart will not turn.
2.
The only sure way to determine which indicators or recorders are good is by observing their operation ani response to transient.
ICS Power Loss 1.
Loss of DC Bus.
a.
All indications on nain control panel are valid except main FW flow arxi neutron error.
These would fail at aid scale.
Al so, same as 2.a below.
Simpson (3/6/80RFI)0!(-91
Response 4 (Continued) 2.
Los s o f "X" AC Power.
a.
Bypa ss val ves a nd fe ed wa ter valves woul d 90 to 50% oper..
Operator would know of this by rapid cooldown rate.
3.
Lo s s of "Y " AC Po we r.
a.
Other than receiving loss of ICS power supply ala rm, operator Juld see no indication of power loss.
The above statements are based on "as-built" conditions prior to the ad-dition of redundant indications which is planned prior to startup of CR-3.
The lists below identify instrumentation that is lost upon loss of NNI(X), NNI(Y), and ICS power.
Also included is a list of the major components required for cold shutdown.
Instruments Lost as a Result of NNI "X" 24 V Power loss TAG NUMBER DESCRIPTION LOCATION MU-14-LT1 MU Tank Level Control Board MU-14-LT2 MU Tank Level Control Board C A-10-T T Boric Acid Temperature All L A LT Boric Acid Level All C A-13-L T Boric Acid Level All CA-14-PT Boric Acid Pump Di scharge All BS-I-dpt 1 RB Spray Flow Control Board BS-3-LT Thiosulphate Tank Level Control Board BS-5-LT Thiosulphate Tank Level Control Board BS-15-PT Thiosulphate Tank Pressure Control Board BS-14-PT Hyd roxide Tank Pressure Control Soard CF-2-LT 1 CF Tank Level Control Board CF-2-LT 3 CF Tank Level Control Board CF-1-PT 1 CF Tank Pressure Control Board CF-1-PT 3 CF Tank Pressure Control Board D.H. 1-dpt 1 DH Flow Control Bcard DH 2 TT 1 DH Ins. Temp.
Compute r MU-27 dpt RC Puap Seal Flow All MU-2 3-d pt 1 HPI Flow Control Board MU-23-dptz HPI Flow Control Board MU-2-PT MU Pump Discharge Pressure Computer MU-17 -PT MU Tank Pressure Control Board MU-24-dpt RC Makeup Flow Control Board MU-31-ft 1 Control Bleedoff Caaputer MU-31-ft 2 Control Bleedoff Computer MU-31-ft 3 Control Bleedoff Camputer MU-32-ft 4 Control Bleedoff Computer MU-7-dpt 1 RC Pump Seal Flow Control Board MU-7-dpt 3 RC Pump Seal Flow Control Board MU-7-dpt 4 RC Pump Seal Flow Control Board Simpso n(3/6/80RF I)DN-94
Instruments Lost as a Result of NNI "X" 24 V Power Loss (Conti nued)
TAG NUMBER DESCRIPTION LOCATION MU-7-dpt 2 RC Punp Seal Flow Control Board SP-7A-dpt Startup FW Flow All SP-8A-dpt 1 Main FW Flow All SP-88-dpt 1 Main FW F1ow Al1 SP-8A-dpt-2 Mai n R! Flow Canpute r SP-5A-TTI FW Temp.
Conputer SP-6A-PT 1 OTSG Pressure All SP-6B-PT 2 OTSG Pressure All SP-10A-PT 2 Turbine Inlet Pressure All SP-108-PT 0 2 Turbine Inlet Pressure All SP-3A-TT I OTSG Downcomer Temp.
All SP-3B-TT I OTSG Downcomer Temp.
All SP-3A-TT 2 OTSG Downccmer Temp.
Co,pute r S P-4 A-TT Mainstream Temp.
All SP-1A-LT 4 OTSG Startup Level All SP-1 A-LT 4 OTSG Startup Level All SP-1A-LT 5 OTSG Startup Level All SP-1B-LT 4 OTSG Startup Level Control Board SP -2B -LT-5 OTSG Startup Level All RC-4A-TT1
" A" Th Remote S/D RC-4B -TT1 "B"
Th Remote S/D SP-1A-LT4 Startup Level Remote S/D SP-6A-PT1 OTSG Outlet Press.
Renote S/D Dti-2-TT1 D.H.
Inj. Temp.
Remote S/D fiU-14-LT1 MU Tank Level Remote S/D MU-14-LT2 MU Tank Level Remote S/D SP-1A-LT 2 OTSG Operate Range Level All SP-1B-LT 2 OTSG Operate Range Level All SP-1A-LT 3 OTSG Operate Range Level All SP-1A-LT 1 OTSG Full Range Level Computer SP-1B-LT 1 OTSG Full Range Level Conpute r RC-10A-PT 1 RC Pump Cavity Seal Pressure Control Board RC-10A-PT 2 RC Pump Cavity Seal Pressure Control Board RC-10B-PT 1 RC Pump Cavity Seal Pressure Control Board RC-10B-PT 2 RC Pump Cavity Seal Pressare Control Board RC-19A-PT 1 RC Pump Cavity Seal Pressure Control Board RC-19A-PT 2 RC Pump Cavity Seal Pressure Control Board RC-19B-PT 1 RC Pump Cavity Seal Pressure Control Board RC-19B-PT 2 RC Pump Cavity Seal Pressure Control Board RC-5A-T T 1 T Na r row Ra nge All c
RC-5B-T T 2 Tc Narrow Range All RC-5A-TT1 T Wide Range All c
RC-5B -T T 2 T Wide Range All c
RC-5A-TT4 T Wide Range Computer c
Simpson(3/6/80RF1)DN-94
Instruments Lost as a Result of NNI "X" 24 V Power Loss (Continued)
TAG NUMBER DESCRIPTION LOCATION RC-4A-TT l Th Tenp.
All RC-4B-T T 1 Th Temp.
All RC-4A-TT4 Th Temp.
Canputer RC-14B-PT RC Flow Control Board RC-1-LT 1 Press. Level All RC-1-LT 2 Press. Level All RC-1-LT 3 Press. Level Control Board RC-2-TT 2 Press. Level All RC-2-TT 2 Press. Level All Instruments Lost as a Result of NNI "Y" 24 V Power Loss TAG NUMBER DESCRIPTION LOCATION SF-1-LT 1 Spent Fuel Level Control Board PF-1-LT 2 Spent Fuel Level Control Board CD-15-PT Condensate Flow Control Board C D-61 -LT Deaerator Level Control Board C A-12-TT Boric Acid Temp.
Waste Di sposal Panel BS-1-dpt 2 BS Spray Flow Control Board B S-7 -T T Thiosul f ate Tank Temp.
Control Board CF-2-LT 2 CF Tank Level All CF-2-LT 4 CF Tank Level Control Board CF PT 2 CF Tank Pressures Control Board CF-1-PT 4 CF Tank Pressures Control Board DH T T 2 DH Inst. Temp.
All DH-6-TT 2 DH Cooler Inlet Temp.
Control Board DH-1-dpt 2 DH Flow Control Board DH-2-TT 1 DH Inst. Temp.
Control Board DH-6-TTl DH Cooler Inlet Tenp.
Control Board l1U-23-dpt 3 HPI Flow Control Board MU-23-dpt 4 HPI Flow Control Board MU-4 -d pt Letdown Flow All M U-5 -TT Letdown Temp.
Control Board MU-2-PT MU Pump Discharge Pressure Control Board MU-18-d pt MU Filter AP Control Board SP-7B-dpt Startup F'.l Fl ow Computer SP-68-PT 1 OTSG Pressure All SP-6A-PT 2 OTSG Pressure Computer SP-10A-PT 1 Turbine Inlet Pressure Conputer SP-1DB-PT 2 Turbine Inlet Pressure Computer SP-4B-TT Mai nstream Temp.
Canputer SP-1 A-LT 2 OTSG Full Range Level Control Room SP-1B-LT 1 OTSG Full Range Level Control Room RC-3A-PT 3 Sa feguards Wide Range Press.
Control Room Simpso n(3/6/80RFI)DN-94
Instrunents Lost as a Result of NNI "Y" 24 V Power Loss
( Conti nued)
TAG NUMBER DESCRIPTION LOCATION RC-3B-PT 3 safeguards Wide Range Press.
Control Room RC-5A-T T 3 Tc Na r row Computer RC-5B-TT3 Tc Narrow Canputer RC-1-LT 3 Pressurizer Level Computer RC -131-PT Low Range RC Pressure All SP-1B-LT4 OTSG Startup Level Computer SP-1B-LT 3 OTSG Oper. Level Control Roon SP-1B-LT4 Startup Level Pemote S/D SP-6B-PT 1 OTS3 Outlet Press.
Remote S/D DH-2-PT2 0.H. In.i. Temp.
Remote S/D Operable / Inoperable Depending on Switch Position SWITCH POSITION TAG NUMBER DESCRIPTION STATUS LOCATION SP-78-d pt Startup FW Flow Bad Control Board N. S.
SP-8A-dpt 2 Mai n FW Fl ow Bad Control Board N.S.
SP-8B-dpt 2 MainFWFlow Bad Control Boarc N.S.
Bad Control Board N.S.
SP-5B-TT 2 Feediater Temp.
Bad Control Board N.S.
SP-5A-PT 2 OTSG Outlet Pressure Bad Control Board N. S.
SP-10A-PT 1 Turbine Inlet Pressure Bad Control Board h.S.
SP-10B-PT 1 Turbine Inlet Pressure Bad Control Board N. S.
SP-3A-TT 2 Downcomer Temp.
Bad Control Board N.S.
SP-3B-TT 2 Downcomer Temp.
Bad All N.S.
SP-18-LT 4 Startup Level Good Co,puter SEL SP-1B-LT 5 Startup Level Bad Ccaputer SEL SP-1B-LT 3 OTSG Level Operate Range Good Control Roo N.S.
SP-1A-LT 3 OTSG Level Oper2+e Range Bad Control Room N. S.
RC-5A-TT 3 Tc Narrow Bad Control Room N.S.
RC-5B-TT 3 Tc Na r row Bad Control Roon SEL RC-5A-TT 4 Tc Wide Good Control Room SEL RC-5B-TT 4 Tc Wide Good All N. S.
RC-4A-TT 4
" A" Th Bad Control Roa, N.S.
RC-48-TT 4 "B"
Th Bad All SP-1B-LT 3 Oper. Level Bad Conputer Tave Ind.
Bad ATC Bad S P-4 A-T T Main Steam Temp.
Bad Control Roou NOTES:
The STATUS shown is for a loss of 'Al "X" 24 V Power Supply for t he swi tc h POSIT 10N s hown.
N.S. - Not Selected SEL - Sel ected Simp so n(3/6/80RFI )DN-94
9 ICS RESPUNSES TO Ntil FAILURES PARAfiETER FAIL POSITION RE SPOU'_
RC F1ow 40 X 106 A multiplier for Loop Btu Limits -
grea tly red uces main FW flow wi t h little time delay due to Main FW Pump Speed Controller.
Total RC Flow 80 X 106 Limits Unit Load Demand (ULD)
Starts a 20%/ min. runback.
Selected Th 570 F A multiplier fo r Loop Btu Limits -
greatly reduces Loop A and B main FW flow.
Selected AT 0F No change - does not alter FW ficu c
ratiet Selected Tave 570 F Tave error modifies reactor dmand -
l ead s to pulling CRDMs to increase power and increase Tave-S/U Level (OTSG) 125" No effect on control of plant.
Operate Level 503 in ef fect on control of plant.
(OTSG)
Steam Pressure 600 Tb ef fect on control of plant.
S/U Flow (FW)
.5 X 106 rio effect initially, but after Block Valve cl o ses, will keep S/U valves o pe n.
Mai n FW Fl ow 3 X 106
.k',a i n FW Fl ow overridden by Stu Limits.
FW Valve AP 50 No ef fect until main FW is less than 50%,
then FW pump speed will i nc rea se.
FW Temperature 350 This error slightly reduces Btu Lim-its and total Fu demarxi signal.
Throttle Press.
900 Turbine valves will open slightly to r egula te pressure inc rea ses MW generated.
Simpso n(3/6/80RF I)DN-94
Major Plant Parameters Needed to Bring Plant to Cold Shutdown Pressurizer Level Pressurizer Temperature Reactor Coolant Loop A Wide Range Cold Leg Tenperature Reactor Coolant Loop B Wide Range Cold Leg Temperature Reactor coolant Loop A Wide Range Hot Leg Temperature Reactor Coolant Loop B Wide Range Hot Leg Temperature Emergency Feedwater Flow A and B Reactor Coolant Systen Wide Range Pressure OTSG A Startup Level 0TSG B Startup Level 0TSG A Operati ng Level OTSG B Operating Level OTSG A Pressure OTSG B Pressure Decay Heat Removal Flow (Both Trains)
Core Flood Tank A Level Core Flood Tank B Level Core Flood Tank A Pressure Core Flood Tank B Pressure High Pressure Injection Flow (Both Trains)
Makeup Tank Level BWST A Level BWST B Level Tsat Meters Letdown Flow Decay Fbat Suction Temperature Low Range RC Pressure Core Exit T/C Indica. tion fiakeup Flow Simpso n(3/6/80RF I)D4-94
REQUEST 5 Address the feasibility of performing a test to verify remaining infor-mation following various Tiill/ICS power upsets.
RESPONSE
Florida Power Corporation will perform a test (s) to verify remaining in-fonnation following various Nil:/ICS power upsets.
Speci fically, we will verify, prior to startup, by f unctional testing that the operator has available the major plant pa r ameters identi fi ed in our res ponse to Request 4 on loss of powe r to NNI(X), IC41(Y ), or ICS.
Addi tio nal det> "
concerni ng these tests will be provided per our discussion of It:
of Correction Action List.
REQUEsis SPECIFIC TO CRYSTAL FIVER UNIT 3 REQUEST 1 Provide a sequence of events for the CR-3 transient on February 26, 1980.
RESPO!!SE to our response to generic Item 3, above, contained the se-quence of events for the CR-3 transient, which occured on February 26, 1980.
REQUEST 2 Provide a list of proposed actions at CR-3 as a result of this event.
RESP 0'lSE Enclosed, as Attachment 2, is our list of corrective actions which will be accanplished prior to restart.
The long-tena action iteas identified in our response to generic Item 3 (Enclosure 5) will be addressed at a later time.
Simpson (3/6/80RFI)DN-94
REQUEST 3 Di scus s the impact, whether it be beneficial or detrimental, of NRC Short-term Lessons Learned and Bulletins and Orders requirements.
RESPONSE
Enclosed, as Attachment 3, is FPC's preliminary assessment of the impact of the Short-term Lessons Learned and Bulletins and Orders requirements.
Simpson(2/6/80RFI)DN-94
~
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y ATTACHf1ENT 1 llistorical Summary of Control Power Failure
$ (1)
DATE IfilTI ATitiG EVErlT RESULT PLAtlT RESP 0ilSE g
07/22/76 Inverter C Trip Loss of Vital Bus C f4one - Pre-opera tional O
3 (2) 08/02/76 Inverter B Trip Loss of Vital Bus B Tbne - Pre-operational
=y (3) 03/02/77 Inverter B Trip Loss of Vital Bus B Reactor Trip from 40% FP - high g
cooldown rate (4) 04/21/77 Blown Fuse in Vital Bus B Loss of ICS 120 VAC Manual Turbine Tri p - fb reac to r trip - 46% FP (5) 08/26/77 Inverter D Trip Loss of Vital Bus D None - Mode 5 (6) 09/02/77 Inverter D Trip Loss of Vital Bus D flone - No reactor trip, 100% FP -
Lost RPS Channel D arti lost some ES - Lost flNI(Y)
(7) 10/26/77 Inverter A Trip Loss of Vital Bus A Reactor trip from 100% FP - Loss of a TIFP (8) 04/17/78 Inverter A Trip Loss of Vital Bus A tbne - Mode 6 (9) 04/21/78 Inverter A Trip Loss of Vital Bus A tbne - Mode 6 (10) 05/04/78 Inverter A Trip Loss of Vital Bus A None - Mode 6 (11) 06/18/78 Inverter A Trip Loss of Vi tal Bus A tbne - flode 6 (12) 02/26/80 ft !I(X) DC Fail ure Loss of flill(X) DC See Sequence of Events f:0T E: ICS is supplial from Vital Bus B llTII(X) is supplied from Vi tal Bus C NI;I(Y) is supplial from Vital Bus 0
ATTACHMENT 2 FPC Corrective Action List Prior to Plant Startup ITEM 1 Thorough testing of the NNI(X) system to determine cause of initial fa il ure.
CLARIFICATION The following five items were checked on the NNI(X) systen to deternine cause of failure.
Procedure PT-454 was developed to examine the systen.
1.
Verified auctioneeri ng diodes and power supply capability to carry bus voltage.
2.
Determined AC voltage level to produce 22 VDC on bus, which trips power supply monitor - 86.8 VAC.
3.
Tested to determine if Si a nd S2 could be closed with load on bus and power supply monitor in service.
S1 a nd S2 time delay
. 5 seco nds.
4.
Interrupted AC power to Si a nd S2 at V8DP to determine power will restore without tripping S1 a nd 5 -
2 5.
Vi sually inspected and tested all Type 820 modules in NNI(X) with 24 VDC supplies.
The first four i tems we re c oupl eted by Ma rc h 2, 1980, and all resul t s were satisfactory.
The following is the results of item 5:
1.
A c onplete mod ul e-by-mod ul e check wa s made of all N'41 ( X )
mod ul e s.
2.
On Wednesday, March 5,1980, a buf fer module was found in the NNI(X) cabinets with the pin connectors touching between the
+24 VDC a nd ground.
The land on the printed circuit board was burned away.
3.
The actual module that fa il ed was installad on Febr ua ry 15, 1980, and was associated wi th the NRC-requi red Psat-T a,} Moni-g tors.
An investigation revealed that the failure in X powe r did not impai r the function of the redundant channel in the "Y" cabinets, nor did any non-NNI Tsat equipnent contribute to the incident.
4.
All buf fer modules in the "Y" cabinets were al so inspected for any problens.
Simpso n(3/6/80RFI )DN-94
ITEM 2 Review PORV circuitry to assure that credible power failures do not cause the PORV to open when it is not requi red to open.
Review power supply independence between PORV and PORV isolation block valve to as-sure that failure which would affect PORV does not eliminate the possi-bility of PORV isolation block action.
CLARIFICATION The PORV circuitry will be modi fied to include an i'terlock from an aux-iliary relay powered from the same source as the
- VDC to the PJHV solenoid.
Loss of either NNI(X) 120 VAC or 24 VDC i l cause the aux-iliary relay to deenergize and block the automatic o, ning circuit of the PORV.
The power supply to the PORV solenoid is 125 VDC, while the block valve is a motor-operated valve, fed fran a 480 VAC Engineered Safety Features motor control center.
Therefore, the power source to the PORV and block valve are independent.
ITEM 3 Modi fy Press uri zer Spray Valve so that NNI power fail ure will close val ve.
CLARIFICATION The Pressurizer Spray Valve will be modified to include an interl ock fran the auxiliary relay described above for the PORV.
Deene rgi zation of this auxiliary relay will cause the automatic opening circuit to be bl oc ked.
The valve will remain in the same position as when the NNI(X) powe r wa s l ost.
The operator will be able to operate the valve froc, the main control board in the remote manual position to either open or close the valve.
Failure of the 120 VAC NNI(X) power or failure of the 24 VJC power to the pressure modules in this instrument control loop will pre-vent the Pressurizer Spray Valve from automatically opening further than previously positioned.
ITEft 4 Provide positive indication of all three (3) Relief Valves.
CLARIFICATION Prior to startup, the modification proposed at CR-3 to provide direct indication of the PORV and safety valves, as described in our submittals of November 17, 1979, January 11, 1980, and February 15, 1980, to Mr. Harold R. Denton, will be installed.
Simpso n(3/6/80RF1)DN-94
ITEM 5 Establish procedur al controls of selectable sources for i ndication and control.
CLARIFICATION Our investigation to-date of the February 26, 1980, transient at CR-3 indicates a need for the following procedural changes which will be im-plemented prior to restart:
1.
Emergency or Abnormal Procedures (Including recognition and response) a.
Loss of vital bus power to a non-nuclear instrumentation
- bus, b.
Loss of vital bus power to integrated control system.
2.
Surveillance Procedures a.
Events Recorder systen.
b.
Instrument systems powe r suppl y and function swi tc h posi tions.
3.
Functional Test Procedures a.
Subcooli ng acnitor.
b.
Redundant instrument avail ability.
ITEM 6 Train all operators and I&C technicians in response to W I and ICS fa il ures.
CLARIFICATIO!!
A traini ng program for all operators and 1&C technicians will be devel-oped and canpleted prior to plant startup.
This training, og= an will i ncl ude:
1.
Review sequence of events for transient of February 26, 1010.
2.
Modifications to NNI power supplies that will be accomplished prior to plant startup.
3.
Detailed description of NNI instrument responses on failure of power sources.
Simpso n(3/6/80RFI)DN-94
4.
Study instrument and plant response on loss of ICS power:
a.
Loss of AC only.
b.
ITEM 7 Move 120 VAC ICS(X) power to Vital Bus.
CLARIFICAT10fl The ICS(X) power supply is presently fed from a regulated power source.
A design to incorporate an Autanatic Bus Transfer ( ABT) between a vital bus powe r source and the regul ated source is being reviewed.
Thi s scheme would orovide backup power to the ICS(X) when either the Vital Bus or the Regulated Instrument Bus is lost.
ITEM 8 Repair Events Recorder System.
CLARIFICATION Analysis of the Events Recorder System following the February 26, 1930, trip of CR-3, revealed that if more than 16 events occurred simultane-ously, everything beyond 16 was lost.
Subsequent troubleshooting iso-lated the cause of this to a defective FS, Sequential Memory module con-necto r problem.
The problem has been corrected and work initiated to i nstall equipment necessary for periodically exerci si ng the f ull capa-bilities of the Sequential Memory circuitry.
ITEM 9 Initiate a more extensive surveillance program on the Events Recorder Systen.
CLARIFICATION A s urve ill ance proc ed ure, SP-50 5, is bei ng devel oped for a periodic functional check of the Events Recorder / Annunciator Systen.
Simp so n(3/6/80RF 1 )DN-94
ITEM 10 Provide operator with redundant indications o' ain plant parameters.
CLARIFICATION The main plant parameter NNI system instrumentation consists of signals o riginated at 120 VAC-powe r ed tra n smi tter s, signal conditioning by 24 VDC-powered electronic modules, and display by 120 VAC-powered i nd i -
cators or recorders.
NNI(X) and NNI(Y) supply all three types of pow-The signal conditioni ng modules are presently mixed between NNI(X) er.
and NN!(Y).
These vital instrument loops will be rewi red to achieve re-d unda nce fran the transmitter through conditioni ng to the i ndica to r.
The indicators will also be rewired to provide 120 VAC power, supplied from the same channel as the signals are dispatched.
A list of the main plant parameters for which the redundant indications will be provided is enclosed in our response to generic item 4.
ITEli 11 Annunciate loss of power to X and Y buses and ICS.
CLARIFICATION There are no present provisions to annunciate the loss of NNI(X) or NNI(Y) 120 VAC and 24 VDC power so urces.
Separate annunciator windows will be provided to infonn the operator of a loss of the NNI(X), the NNI(Y), or ICS pover sources.
ITEM 12 Devel op a nd insti tute a test pr og ram fo r changes in designs and modi fica tio ns.
CLARIFICATION Functional testi ng of modified portions of the NNI systen will be accom-pli shed in accordance with approved plant procedures.
The developnent and conduct of such test procedures is governed by existing Compliance Procedure CP-114, " Procedure for Preparation and Co ntrol of Permanent Modifications, Temporary Modi fications, Deviations, and MAR Functional Te st Proced ures".
Specific test procedures will be devel > ped, issued, and approved as part of the construction work package authorizing and detailing the installa-tion once the systen modifications / redesigns have been finalized.
Simpson(3/6/80RFI)DN-94
ITEM 13 Install indicating 1ights on all vital bus feeds.
CLARIFICATION An indicati ng light for each branch circuit coming fran the safety-re-lated vital distribution panels will be provided.
The indicating lamps and i solating f uses will be installed in a box which will be mounted near the distribution panels.
There is a separate box for each distri-bution panel.
ITEM 14 Modify vital bus panels for quick fuse replacement.
CLARIFICATION The vital bus panels will be modified to facilitate access to each panel to permit quick fuse replacement.
ITEM 15 Modify EF Pump auto start circuit and reactor trip circuit so that any power fail ure will not pr event activation on low SG level (control g rad e).
CLARIFICATION The present control grade emergency feedwater pump automatic start cir-cuits and reactor trip ciruits will not actuate if NNI(X) power is lost to the l ow-l ow steam generato r l evel sig nal monitors.
A red undant NN!(X) and NNI(Y) low-low steam generator level signal fror, each steam generator will be imp'.emented such that loss o' either NNI(X) or NNI(Y) 120 VAC or 24 VDC power will not prevent the actuation of the control grade emergency feedwater pump automatic start circuit and reactor trip ci rcui t.
ITEM 16 PORV valve position indicating lights for solenoid will be added.
CLARIFICATION PORV valve position indicating lights indicate selector switch position, high pressure setpoint select, and low pressure setpoint select.
These PORV i ndica ti ng lights will be modi fi ed to provide more meani ngf ul Simpson(3/6/80RFI)0N-94
information to the operator.
In addition, a new systen is presently be-ing installed to provide a more positive PORV valve indication to the o pe r ato r.
ITEM 17 Visually inspect the lower portion of the steam generator support skirts and anchor bolts.
Remove any corrosive residue observed.
CLARIFICATION This will be completed prior to restart of CR-3.
ITEM 18 Inspect the pressurizer heater bundles for seal leakage.
Electrically check the pressurizer heater elements for continuity.
CLARIFICATIO" This will be accomplished prior to restart of CR-3.
ITEfi 19 A fatigue analysis will be performed on the pressurizer heaters to dem-onstrate that the 40-year design life was not adversely impacted by this transient.
CLARIFICATION FPC will have B&W perform this analysis prior to the restart of CR-3.
ITEM 20 The relief valve loadings on the pressurizer relief nozzles should be detennined and their effect will be assessed.
CLARIFICATION FPC will have B&W, with the assistance of GAI, perfona this assessment prior to the restart of CH-3.
Simpson(3/6/80RFI )DN-94
ITEM 21 The CRDMs will be checked for proper i ns ula tion resi stance prior to their return to service.
CLARIFICATION This item will be coapleted prior to the restart of CR-3.
ITEM 22 a.
Perform a visual inspection of the pressurizer relief system (i e.,
PORV, both code safety valves, and the discharge piping).
Ins pec-tion of the discharge piping system, including hangers, should be performed to ensure that no gross distortions have occurred, b.
Confi nn by calculation that the structural loads imposed on the valves and pressurizer as a result of the extended period of di s-charge to the quench tanks are acceptable, c.
Di sa s s embly, i ns pection, a nd ref urbi shnent (as necessary; of the PORV and the code safety valves.
CLARIFICATION Iten 22.a a nd c wil l be acconplished by FPC prior to the restart of CR-3.
J FPC has requested the assistance of GAI and B&W to verify that the loads imposed on the valves and pressurizer duri ng the transient are accepta-ble.
Thi s i tem wi l l be t an pl eted pr i o r to sta r t up of CR-3.
ITEM 23 Provide diverse containment isolation.
CLARIFICATION The diverse containnent isolation modification proposed in response to Short-term Lessons Learned Item 2.1.4 will be installed prior to startup of CR-3.
This modification is described in our response dated Febru-ary 15, 1980, to Mr. Harold R. Denton.
Simpso n(3/6/80RF I)Dil-94
ATTACHMENT 3 NRC Short-Tena Lessons Learned and Bulletins and Orders Review Summary Based on a quick review of NRC Short-Tenn Lessons Learned ar.d Bulletins and Orders Requirements.
Florida Power Corporation has determined the overall impact of these requirements.
The impact is briefly sumuarized from a high level management overview.
The significant benefit of im-plerenting Short-tena Lessons Learned, Bulletins and Orders, and FPC in-ternal requirements following TMI-2 was the aid to the plant staff in minimizing the severity of this incident.
Due to the relative minimal severity of the incident and inability to determine the errors which were precluded, the overall benefit or detriment of the requirements cannot be fully quantified.
We will continue our review of this impact anc provide additional information prior to restart based on a more th) rough in-depth review.
Training - The most sicnificant benefit as a result of the NRC Short-Term Lessons Learned and Bulletins and Orders, is in the area of train-i ng for plant operators.
Specifically, emphasis in training relating to the following areas enabled the plant operators to maintain the plant in a stable condition at all times with no adverse impact to plant systems and/or equi pment:
- PORV isolation
- Maintaining adequate Subcooling Margin
- Contai nnent Isolation
- Tripping of the Reactor Coolant Pumps
- Emergency Feedwater Initiation
- High Pressure Ignition (HPI) Actuation and Operation
- Small Breaks / Inadequate Core Cooling Operating Procedures
- Shif t Supervisor Responsibilities
- Shif t Technical Advisor's Responsibilities
- Shift Turnover Procedures
- Control Room Access
- Emergency Operating Procedures
- futural Circulation Equipnent and Facilities - Based oa our preliminary review of the equip-Fent anfficiTity modTricetions which have been implemented as a result of NRC Short-Tena Lessons Learned and Bulletins and Orders Require.nents.
the only nodifications that had any impact were the Tsat Meter, Incore Tempe, ature monitor, Post-Accident Sampling Capability, Emergency Feed-water Flow, Containment Isolation, the Technical Support Center, and Operational Support Center.
Tsat Meter -
This instrumentation allowed the operator to verify an adequate subCooling nurgin throughout the transient.
This is Simpson(3/6/80RFI)DN-94
one of the parameters used by the operator to determine the need for HPI flow requir ements.
Incore Temperature Monitor - This instrumentation was used by the operator to verify by manual calculations that adequate subcooling margin was being maintained.
Post-Accident Sampli ng Capability - The sampli ng capability instal-l ed fo r po st-accid ent sampling was utilized during this incident.
However, dte to the relative low levels of activity in the reactor coolant, our previous sampling capability and procedures would have been sufficient for this transient.
Emergency Feedwater Flow - This instrumentation was availa31e to the operator but had little impact on the event.
Contai nment Isolation - As a result of our responses to bu,,
. ins related to Dil-2, a short-term instruction was issued at CR-3 which isolated nonessential systems that containad or could contain radi-oactive ef fl uents.
Incl ud ed in this was the waste gas systen to the RC Drain Tank.
This caused the rupture disc to relieve earlier in the transient than it would have prior to this administrative r equi rement. This was due to the buildup of waste gas in the drain tank prior to the event.
However, this administrative action did preclude the transfer of radioactive effluents outside of the Con-tai nnent Buildi ng duri ng the event.
Technic al Support Center and Operational Support Cente r These were manned duri ng the transient a ni provid ed a foc al point for feedback of communications, personnel assembly point, and reduction in the number of personnel in the Control Room.
However, due to the short duration and relative minimal severity of the transient, their contributions were minimal.
In addition, the TSC was util-ized as the point of contact with the NRC, in conjunction d th the Control Roon.
_ Conclusion - Our review to-date of the NRC requirenents imposed si nc e Dil-2, has concluded that these requirenents were of benefit to plant o pe r atio ns.
Operator training since TMI-2 provided the largest contri-bution for deali ng with this transient.
Simpson(3/6/80RFI)DN-94
Rev. 5 ENCLOSURE 1 F3E' 1 SEQL'ENCE (AS OF 2300 3/1/S0) 26 February Transient CR-3 EVENT SYNOPSIS At 14:23 on February 26, 1980 Crystal River -3 Nuclear Station experienced a reactor trip from approximately 100% full power. A synepsis of key events anc parameters was obtained from the plant ce=puter's post-trip review and plant alarm su==ary, the sequence of events monitor, control room strip charts, and the Shif t Supervisor's log.
The reactor >was operating at approxiestely 100 P. full power with Integrated Centrol System (ICS) in aute=atic. No tests were in progress and =iner main-tenance was being perforced in the Non-Nuclear Instru=entatien (NNI) cabinet "Y".
Ti=e Event Cause/ Cec =ents 14:23:00 The following is a summary of plant conditions prior to tha trip Flux 98.6%
RC Pressure 2157 psig PZR level 202 inches MU tank level 71 inches T
"A" 599'F.
TN "B" 600*F.
TH "A" 557*F.
T "B" 556*F.
RbFlcw"A"73X106 lbs/hr RC Flow "B" 73 X 106 lbs/hr Letdown Flow 48 gpm OTSG "A" lvl (0P) 677.
OTSG "B" lvl (0P) 65 OTSG "A" FRLV 242 inches OTSG "B" FRLV 254 inches OTSG "A" Pressure 911 psig OTSG "B" pressure 909 psig Main Steam Pressure 894 psig Main Steam Temp. 569 F.
Condenser Vacuum 1.76 Generated MW 834 DFT level 12.7 ft.
6 Feed. Flow "A" 5 X 10 lbs/hr Feed Flow "B" 5 X 106 lbs/hr Feed Pressure "A" 970 psig Feed Pressure "B" 968 psig 14:23:21
+24 Volt Bus Failure (NNI Cause still unknown. Apparenti.,
power loss "X" supply) the positive 24 VDC bus shorte.i dragging the bus voltage dcwn te a
Rev. 5 Page 2 Ti=e Event Cause/Cc :ents low voltage trip cenditier..-
is a built-in k to h sec:nd
- c__ -
at which time all power suprl_es will trip. There was no trip indication on negative (-) ic':sge.
This event was missed by the annunciator. Follcwing the NN1 pcwer failure, much of the centrol ::c:
indication was lost. Of tha S n:trJ~-
entation that renained c; erat's transient conditions made the"
'ndic-cation questionable to tne c:e-ators.
14:23:21 PORV and Spray Open When the positive 24 VOC suppit was lost due to the sequence discussed abe; the signal moniters in N'.I changed sta:
causing PORV/ Spray valves to open. The PORV circuitry is designed to seal in upon actuation and did so. The resul:an loss of the negative 24 VDC halted spray valve noter operater and prevente PORV seal in from clearing en icw pressure. It is postulated tha: the PCR opened fully and the spray valve str:ke for approximately b secend. The AC'; cpe indication on spray valve did net actuate, therefore, the spray talve did not exceed 40" open.
14:23:21 Reduction in Feedwater As a result of the "X" power suppl:
failure many primary plant centrol signals responded erronecusly. Tceld failed to 570*F (nor:al indica:icn was 557*F) producing several spuricus alarn Tave failed to 570* F (decreased'-
The resultant Tave error modified :'.e react de=and such that control rods were Withdrawn
.to increse Tave and reac: :
power. The 'pewer increase was t ernin i.c at 103" by the ICS and a " React:- Demar High Limit" alarn was received. Thet failed to 570* F (lcw) and RC fl.. "aile 6
to 40 X 10 lbs/hr in cach icer Gew}.
Both these failures created a E!! alarr and limit on feedwater which reduced feedwater flow to both OTSG's te S
essentially zero. Turbine Header
?sst f ailed to 900 psig (high) which cruced the turbine valves to open slight te
Re.;
5 Page 3
_Ti=e Event Cause/Cc==ents regulate header pressure thus incre: sing generated megawatts. These ec=bined failures resulted in a loss of heat sink :: the reactc initiating an excessively high RC pressure condition.
14:23:35 Reactor Trip / Turbine Rx trip caused by high RCS pressure at '.300 ;
Trip Turbine was tripped by the reacter.
14:24:02 Hi Pressure Inj.
This was a ec=puter printout and indi:stes Req. (Flag)
<50* subcooling.* See attached graph of RC Pressure /Te=p. vs. Time. This graph is based on Post Trip data and actual intore ther:c-couple data. From the reactor trip point (la to 14:33, core exit te=perature data was obtained by extrapolatica and calculated dat This is supported by two alar: data points plotted at 18* and 21* of subcecling during this period frem the ce=puter. It is i= porta-to note that lcwest level of subecoling was 8'F for a very short period of tire.
- NOTE: This ec puter program was initiated a a result of the TMI incident.
14:24:02 Loss of Both Suspect condensate pump tripped due te high Condensate Pu=ps DFT level. This is verified by ' printed by computer, indicating the level instrument was over ranged as well as a lew flew indication in the gland steam condenser as a indicated by cc=puter.
14:25:50 PORV Isolated At this ti=e a high RC Drain Tank level alar was received. This was resultant frca the PORV re=aining open and was positive indica:
that the PCRV was open. At this tire, the operator closed the PORV block valve due to RCS pressure decreasing and high RCTT level 14:26:41 HPI Auto Initiation HPI initiated autccatically due to ic. RCS pressure of 1500 psig. The icw pressere conditien was resultant from the PCK'; re:ain full open while the plant was tripped. Full HPI was initiated with 3 pumps resulting in approximately 1100 gpm flow to the RCf. At this time, all remaining non-essentii-R.R.
isolation valves
Rev, o.
Page -
Ti=e Event Cause/Cc==ents were closed per TMI Lessons Learne: {utdeline 14:26:54 RC Pu=ps Shutdown Operator turned RC pu=ps of f as recuired by the applicable e=ergency procedure and 3 5 'i small break guidelines.
14:27:20 RB Pressure Increasing This is first indication that RCDT rupture disc had ruptured.
R3 pressure increase datr was obtained fro = Post Trip Review and Strip Cha:
indication.
14:31:32 R3 Pressure High This alarm was initiated by 2 psi; in R3.
T!
is attributed to steam release frc: RCDT.
Ci safeties had not opened at this ti=e based u; tail pipe te=peratures recorded at 14: 32:03 (Computer).
14: 31:49 OSTG "A" Rupture Matrix This occurred due to <600 psig in OTSS "A.
Actuation The low pressure was caused by OTSG "A" boil dry which was resultant from the 3-"
a-failed OTSG level trans=itter.
This resulte in the closure of all feedwater and steam bl.
valves which service OTSG "A".
14:31:59 Main Feedwater Pu=p 1A Caused by suction valve shutting due to Tripped
= atrix actuation in previous step.
14: 32+14:41 ES A/3 Bypass Manually bypassed and HPI balanced betwcer a
4 no :les (Total flow approximately 1120 gp-
-scall break operating guidelinesT.
14:32:35 Started Stea= Driven Started by operator to ensure feedwater ens Emergency Feedwater Pump available to feed OTSG's.
14:33 Core Exit Te=p. Verified The core exit incore ther= occupies indica t ed the highest core outlet te=rerature value 560*F.
RCS pressure was 2353 psig atthis ti:
therefore, the subcooling margin at this t ir-was 100*F.
Mini =u: subcooling =ar;in far th.
entire transient was 8 F.
It is postulated that some localized beiling occurred in the core at this peint as indicated by the self powered neutr-detectors.
14:33+14:44 Started Motor Driven E=er-Sa=e discussion as " Started Stea: Oriren E=e:
gency Feedwater Pump gency Feedwater Pu=p."
14:33:30 RC Pressure High (2395 psig) At this point, pressurizer is solid and cede safety lifts (RCV-8).
This is the highest RCS pressure as recorded on Post Trip Review Apparently, RCV-8 lifted early due to seat
Rev. 5 Page 5 Time Event Cause/Cor=ents leakage prior to the transient and RC;-; did not lift.
14:34:23 RB Do=e Hi Rad Level RMG-19 alarmed at this point. Highest lerel indicated during course of incident was 50 R/hr. High radiation levels in RB caused by release of non-condensable gases in the press urizer and coolant.
14:35:33 Attempted NNI Repower With-This resu. 2d in spikes observed on de-ener-out Success gized strip charts.
14:36:50 Computer Overload Caused by overload of buffer.
Resulting in no further computer data until buffer catches up with printout.
14:38:15 FWV-34 Closed This valve was closed to prevent overfeeding OTSG "B" beyond 100% indicated Cperating Ran-.
14:44:12 NNI Power Restored Success-NNI was restored by removing the X-N'C Pcver fully Supply Monitor Module.
This allowed the breakers to be reclosed. At this ti e, it wi observed that the "A" OTSG was dry, the pres:
urizer was solid (Indicated off scale high),
RC outlet temperature indicated 556*F (Loop.
& B average), and RC average te:perature ind.
cated 532*F (Loop A & B).
The highest core i thermocouple temperature at this time was 53 RSC pressure was 2400 psig (saturation terp.
this pressure is 662*F.).
This data cerific.
natural circulation was in oregress api the plant subcooling margin was 131*?
(base en core exit the rmocouples).
14:44:31 RB Isolation and Cooling Actuation At this time, RB pressure increased tc 4 pr %
and initiated RB Isolation.
The oper.te-verified all immediate actions occurre-prep.
for HPI, LPI, and R3 Isolation and Ceci ac.
increasing RB pressure was resultant frc RC:
par, sing HPI at this ti=c.
14:46:10 Bypassed HPI, LPI and RB These "ES" systems were bypassea at :
1.<
t iri Isolation and Cooling to again balance HPI flow and restere e c o li n.
water ssential auxiliary equipmen (i.e.
6.
RCP's, letdown coolers, CRDM's etc.T.
Rev. 5 Page 6 Time Event Cause/ Con =ents 14:51:57 Rupture Matrix Actuation on The actuation was resultant fro = a de;-
OTSG-B radation of OTSG-B pressure. Cold e=er-gency feed was being injected in:o the OTSG at this ti=e.
This = atrix actuation : scla:ed all feedwater and stea= block valves to the B-0TSG and tripped the "B" = sin FW p =;- 3oth E=ergency FW pu=ps were already in c;cration at this ti=e.
B-0TSG level at this time was 70% (Operation Range).
14:52 HPI Throttled and RCS At this ti=e, the =axi=u: core exit thcr=c-Pressure Reduced to 2300 couple te=7erature was 515'F, RCS pressure psig was 2390 psig.
Therefore, the subcooling nstgin was 147'F. Natural circulation was in effect as verified previcusly. All cen-ditions had been satisfied to thrc::le EPl.
Therefore, flow was throttled down :: approx-imately 250 gp= to reduce RCS pressure te 2300 psig in order to attempt to reduce the flow rate through RCV-8 and into the R3.
14:53 Reestablished Letdown At this ti=e, the operator was a::e ptin; to establish RCS pressure control via ncr=a2 RC =akeup and letdown.
14:56 Opened MU Pu=p Recire.
This was done to assure the MU pu=ps wculd Valves have mini =u flow at all times to prevent possible pump da= age.
14:56:43 Bypassed the A-0TSG Rupture Feedwater was slowly ad=i: ed Matrix and Reestablished to the A-0TSG which was dry up to this pair:-
Feed to the A-0TSG Feedwater was ad=itted through the Auxiliar:
FW header via the EFW bypass valves.
The feedrate was very slow in order to =inicine ther=al shock to the OTSG and resultant depre surization of the RCS.
RCS pressure centre!
was very unstable at this ti=e.
It is pesi.--
that some localized'boilint occured t-tere at this point as indicated bv self ne trer detectors.
Rev. 3 Page 7 Ti=e Event Cause/Cecrents 14:57:09 Bypassed the B-0TSG This was done to regain FW centre'.: the Rupture Matris B-0TSG. Level was still high in this C!SG (approx 1=ately 65% Operating Rance). Therefore feed was not necessary at this tire. The Main Steam Isolation valves were open in preparation for bypass valve operati:n (when necessary).
14:57:15 Established RC Pu=p This was done in preparatien for a Fi? start Seal Return (when necessary) and to minimize pur; sea'_
degradation.
15:00:09 Reestablished Level This verified feedwater was being 2dritta: t In A-0TSG the OTSG and made it available fer core cooling via natural circulation. Feed tc this generator was continued ith the in ent of proceeding to 95% cn the Operating Rat. e.
15:00:09 77'F Subcooled "A" Loop This value was based upon "A" RCS Icop parameters at this time. The "A" loop was being cooled down at this time by the A-CTSG fill and the operator was attenpring te equalize loop temperatures.
15:15 23*F Delta-T/ Manned the At this ti=e, loop te peratures were nearing Technical Support Center equalization. This delta-T was calculated from loop A & B T 's and core exit ther:c-c couples.
15:17 Declared Class "B" E=ergency This was done based en the fact there was a loss of coolant through RCV-3 in the contain=ent and HPI had been initiate. All non-essential CRd 3 personnel were directed to evacuate and. contact off-site agencies 'e gan.
Survey team was sent to Auxiliar, 3u '.c 15:19 Opened E=ergency FW Block At this point the A-0TSG level was increscing to B-0TSG and the decision was made to commence fill the B-0TSG simultaneously. The intent was ts go 95% on both OTSC's without exceedin-FC5 cooldown limits (107 F/hr) while maintainin:
RCS pressure control.
t
EeV. f Page i Time Event Cause/Co cents 15:26 Lo Level Alarm in Sodium This was recultant from the tank suppi;. valve Hydroxide Tank opening when the 4 psig R3 isolation and cool ing signal actuated.
The sodiu= 'f r:r.ide wa d
released to both LPI trains.
Sediu: u d::xid.
was ad=itted to the RCS via HP1 ft:: the I'45!
('Approxi'=ately 2.ppe inj ected int o the ?.:5.)
15:50 Terminated HPI At this time, all conditions had been satis-fied (per s=all break operating gutdelines) to terminate HPI.
RCS pressure control had been established using nor=al takeup and letdown. EPI was terminiated and essentially all releases to the R3 were discontinued.
16:00 Co=menced Pressuri er At this ti=e, RCS pressure and te=perature Heatup were well under control. Natural circulatier was functioning as designed (approximately 22 delta-T). RCS temperature was being raintair at approxi=ately 450*.
RCS pressure was app:
imately 2300 psig.
The decision was made at this point to co==ence pressurizer heatup in preparation to re-establish a stea= space in the pressurizer.
16:07 Survey Team Report The E=ergency Survey Tea = reported no radiat:
survey results taken off site were above back-ground.
16:08 :04 Shutdown Steam Drive E=ergency FW Pump The =otor driven E=ergency FW pump was runni:
therefore, the steam driven pu=p was not nee; The plant re=ained in this condition fcr app-roximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, while heating up the pres urizer to saturation temperature for 15C0 ps:
16:15 Press Release Media was notified of plant status.
18:05 Established Steam Space Pressurizer At this point, pressuriser te=perature was approximately 620*F.
Pressurizer level was brought back on scale by increasing le:down.
From this point pressurizer level was reduced to normal operating level and normal rressure was established via pressure heaters.
18:30 Terminated Class B E=ergency State and Federal Agencies notified.
Rev 3 Pagt 9 Ti=e Event Cause/Co==ents 21:07 Forced Flow Initiated The decision was made to re-estah'ish for:ed in RCS flow cooling in the RCS at tnis ti_e.
35W and NRC were consulted. RCP-13 and 10 were started. At this point, RCS para:eters were stabilized and =aintained at RC pressure-200C psig, RCS temperature 420*F.
Pressurizer level-235 inches.
The plant was considered i a nor=al configuration.
6
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I l } i l-CR 3 - SEQtEMEE OF E\\ENTS (CONT'D> l THE ETtT f 15:17 DECLARED CLASS *B" EtEfGENCY l 15:50 TER11NATED HPl 16:00 00HENCED PRESSLRIZER HEATLP f 16:07 StR\\EY TEA 1 REPORT 16:08:04 SHlEChN STEA1 DRI',EN EtEfGENCY Pd PliP 16:15 PRESS RELEASE 18:05 ESTABLISHED STEA1 SPACE IN PRESSLRIZER -l 18:30 TERilNATED CLASS *B" EtERiENCY 21:07 FORCED FLOW INITIATED IN RCS 1 l i l i 1 i t i f i I
I i 2_1 SECLENCE OF EtE1TS Tf1E E\\ENTS 14:23:00 LNIT AT 98:.6% POWER 14:23:21 +24 \\0LT B LS. FAlLBE (NNI' POWER LOSS) 14:23:21 PORV ANDi SPPAY OPEN 14:23:21 REDtETION IN FEEDWATER 14:23:35 REACTOR TRIP /TLRBINE TRIP 14:24:02 H1 PRESSLRE INJ. REQ. (FLG ) 14:25:50 PCRV ISOLATED 14:26:41 HPI ALTO INITIATICN 14:26:54 RC PLMPS SHLTDOWN 14:27:20 RB PRESSLRE INCREASI.'E i. 14:31:49 OTSi "A" RlPTLRE i ATRN j 14:32:10:44 STARTED BOTH E1E?iENCY FEEDWATER PL'1PS { 14:33 CORE E(lT TEiP. \\ERIFIED i 14:33:30 RC PRESSlRE HriH (2395 PSri) 1, 14:34:23 R3 D01E HI RAD LE',EL f 14:44:12 NNI PCWER RESTORED SLCCESSflLLY f 14:44:31 R3 ISCLATION AND C00LINi ACTLATION l 14:46:10 BYPASSED HPI, LPI, AND RB ISGtATICH AND C00LINi l 14:51:57 RlPTLRE 1ATRI( ACTLATION ON OTSi-B j 14:52 HPI THROTTLED AND RCS PRESSLRE REDLEED TO 2300 PSli. 8 14:56:43 BYPASSED THE A-0TSi RlPTlRE 1ATRN AND RE-ESTABLISHED FEED TO THE A-0TS'i 14:57:09 BYPASSED THE B-0TSi RlPTLRE 1 ATRN { 15:00:09 RE-ESTABLISHED LEtEL IN A-0TSi 15:15 1ANNED THE TECHNICAL SLPPORT CENTER i
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ENCLOSURE 3 DESCRIPTION OF EOUIPMINT DEFICIENCIES The following equipment deficiencies were noted during an in depth review of the. Crystal River Unit 3 transient on February 26, 1980. 1. Loss of the Non Nuclear Instrumentation (NNI) System (X) DC power supplies. 2. Power operated relief valve (PORV) failed open. 3. Pressuri:er spray valve failed open. 4. Loss of indication of =ajor plant parameters. 5. Integrated Control Syste= (ICS) response to NNI (X) failure. A. Auto initiation of Emergency Feedwater. B. Reactor trip on loss of feedwater. 6. ICS loss of 120 v AC for 2 seconds. 7. Events recorder degradation. 8. Computer alarm buffer overflow. 9. Systems that perfor=ed adequately. A. Reactor Protective System. B. Engineered Safeguards Activation. C. Control Rod Drive System. D. Steam Rupture Matrix. E. Secondary Plant Equipment. F. Radiation Monitoring System.
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ENCLUSURE 5 Corrective Action Prior to Plant Startu:i 1. Thorough testing of the NNI (X) systen to determine cause of ini-tia. fail ure. 2. Review PORV circuitry to assure that credible power failures do not cause the PORV to open when it i s not requi r ed to open. Review powe r supply independence between POR V a nd PORV isolation bloci valve to assure that a ny si ngle fail ure which wo uld affect f' #.' does not eliminate the possibility of PORV isolation block action. 3. Modify Pressurizer spray valve so that Niil power failure will closo val ve. 4. Provide positive indication of all three (3) Relief Valves. 5. Establish proced ural controls of selectable sources for i nd ic ati o n a nd control. 6. Train all operators and I&C technicians in response to TGI and ICS fa il ur es. 7. Move 120VAC ICS(X) power to Vital Bus. 8. Repair Event s Recorder System. 9. Initiate a nore extensive surveillance progra, on the Events Recar-der Syster. l '). Provide operator wi th redundant i nd ic a ti ons of main pl a nt parameters. 11. Annunciate loss of power to X and Y buses and ICS. 12. Devel op a nd institute a test progra, for cha n3es in desi;ns anf nodifications. 13. Install indicati ry light on all Vital Bus feeds. 14. Ilodi fy Vi tal Bus panel s fo r quick fuse replacement. 15. Modify EF Pump auto start circuit and Reactor trip circuit se that any power failure will not prevent activaticn on low Sti l ev el (con-trol grade). 16. PORV valve position indicating lights for solenoid will be added. 17. Visually inspect the lower portion of the stean generator suppori ski r ts and anchor bol ts. Renove any corrosive residue obset v6!. Simpso n(3/6/80RF 1 )Dil-94 J:(U a
18. Inspect the heater bundles on the pressurizer for seal leakage. Electrically check t he hea ter el ement s on t he pressurizer for conti nui ty.
- 19. A fatigue analysis will be perfo rmed on the pressurizer heaters to demonstrate that the 40-year design life was not adversely impacted by this transient.
20. The relief valve loadings on the pressurizer relief nozzles should be determined and their ef fect will be assessed. 21. The CRDMs vdll be checked for proper insulation resistance prior to their return to service. 22. The following will be accomplished prior to restart: a. Per fo rm a vis ual inspection of the pressurizer relief systen (i.e., PORV, both code safety valves, and the discharge pip-i ng). Ins pection of the di scharge pi pi ng systen including hangers should be pe r fo rmed to ensure that no gross distor-tions have occurred. b. Confi rm by calculation that the axial loads and moments im-posed on the valves and pressurizer as a result of the extend-ed period of discharge to the quench tanks are no creater than those considered in the original design evaluation. c. Disassembly, inspection and refurbishment (as required) of the code safety valves. 23. Install diverse containment isolation modification. Simpson(3/6/80RFI)DN-94}}