Skip to Main Page ContentSkip to SearchSkip to Site Map NavigationSkip to Footer Links

        - Home

        - FAQ
        - Glossary
        - Facility Locator
        - What's New
        - Site Help
        - Index A-Z
        - Contact Us
        - Email Updates
        -    Listen to page

Report a Safety Concern

      - Nuclear Reactors  - Power Reactors
- Research & Test Reactors
- Operating Reactors
- Operator Licensing
- New Reactors
- Operator Licensing for New Reactors
- Research Activities
- Nuclear Reactor Quick Links
 
      - Nuclear Materials  - Special Nuclear Material
- Source Material
- Byproduct Material
- Med, Ind, & Academic Uses
- Radium
- Uranium Recovery
- Fuel Cycle Facilities
- Materials Transportation
- Research Activities
- Nuclear Materials Quick Links
 
      - Radioactive Waste  - Decommissioning of Nuclear Facilities
- Low-Level Waste
- Waste Incidental to Reprocessing
- High-Level Waste
- Uranium Mill Tailings
- Low-Level Waste Disposal
- High-Level Waste Disposal
- Storage of Spent Nuclear Fuel
- Transportation of Spent Nuclear Fuel
- Research Activities
- Radioactive Waste Quick Links
 
      - Nuclear Security  - Domestic Safeguards
- Information Security
- Radioactive Material Security
- Required Reporting for Clearance Holders
- Insider Threat Program for Licensees
- Criminal History & Firearms Checks
- Contact Us
 
      - Public Meetings & Involvement  - The NRC Approach to Open Government
- About Meetings Open to the Public
- Conferences & Symposia
- Documents for Comment
- NRC Rules and Petitions
- Facilitating Stakeholder Involvement
- NRC Information Quality Guidelines
- Subscribe to E-mail Updates
- Commission Schedule
- Adjudications (Hearings)
- Public Meeting Schedule
- Adjudicatory Submissions
 
      - NRC Library  - Basic References
- Document Collections
- ADAMS Public Documents
- Public Document Room
- LSN Library
- FOIA & Privacy Act
- Records Management
- Training Courses
- FAQ Index
- Get Copies of Documents
- Withholding of Sensitive Information
- Photos & Video
- Electronic Hearing Docket
 
      - About NRC  - The Commission
- Organization & Functions
- Governing Legislation
- Plans, Budget, & Performance
- Locations
- History
- Values
- Direction-Setting & Policymaking
- Research Activities
- Radiation Protection
- Fire Protection
- Safety Culture
- How We Regulate
- Emergency Preparedness & Response
- Public Affairs
- Congressional Affairs
- Enforcement
- International Programs
- State & Tribal Programs
- Alternative Dispute Resolution Programs
- Privacy Program
- Civil Rights
- Contact Us
- Career Opportunities
- Contracting Opportunities
- Small Businesses
- Grant Opportunities
 

Print

HomeNRC LibraryDocument CollectionsGeneric CommunicationsInformation Notices1980

# Information Notice No.80-10 – Partial Loss of Non-Nuclear Instrument System Power Supply During Operation

```
                                                            SSINS No.: 6870 
                                                            Accession No.: 
                                                            8002280640 

                                UNITED STATES
                        NUCLEAR REGULATORY COMMISSION
                    OFFICE OF INSPECTION AND ENFORCEMENT 
                           WASHINGTON, D.C. 20555 

                               March 7, 1980 

                                           Information Notice No. 80-10 

PARTIAL LOSS OF NON-NUCLEAR INSTRUMENT SYSTEM POWER SUPPLY DURING OPERATION

Description of Circumstances: 

This notice contains information regarding Crystal River Unit 3 response to 
a loss of non-nuclear instrumentation (NNI) as a consequence of loss of the 
+24 volt power supply to the NNI.

At 2:23 p.m. on February 26 with Crystal River Unit 3 at 100% power, the +24 
volt power supply to the NNI was lost, due to a short to ground. This 
initiated a sequence of events (detailed in the enclosure) wherein the PORV
opened and stayed open as a direct result of the NNI power supply loss. HPI
initiated as a result of depressurization through the open PORV, and with
approximately 70% of NNI inoperable or inaccurate, the operator correctly
decided that there was insufficient information available to justify
terminating HPI. Therefore, the pressurizer was pumped solid, one safety
valve lifted, and flow through the safety valve was sufficient to rupture 
the RC Drain Tank rupture disk, spilling approximately forty-three thousand 
gallons of primary water into containment. 

The Crystal River 3 event is closely related to the November 10, 1979 event
at Oconee Unit 3 wherein the inverter supplying power to the Integrated
Control System (ICS) and to parts of the NNI failed. That event was the
subject of Information Notice No. 79-29 (November 16, 1979) which was 
followed by IE Bulletin 79-27 (November 30, 1979). 

The CR-3 event involved loss of only part of the power available from an
inverter, rather than the inverter itself, since the +24v supply is only one 
of several power supplies drawing power from one inverter. The effects are 
very similar, however, in that the ICS lost part of its input signals in both
events.

The +24 volt power supply short to ground has tentatively been identified by 
the licensee to have occurred between knife edge connectors of a Bailey 
Control Company Voltage Buffer Card. The voltage buffer card was misaligned
in its receptacle, and adjacent connectors carrying +24v and "common" were
bent such that they contacted one another. This short circuit cleared itself
during subsequent re-energizing of the power supply by burning through the
foil on a printed circuit card. Subsequent review by the licensee identified
a second voltage, buffer card which was also misaligned but had not caused
a short circuit. The specific circuit cards which were misaligned carried
part number 6624609L1. The connectors on these cards are slightly thinner 
and appear to have a somewhat different angle than those found on similar 
cards elsewhere in the NNI which carry part numbers 6624608A1 or 6624609A1. 
The 6624609LI cards appear to be more subject to misalignment. 
.

Information Notice No. 80-10                             March 7, 1980  
                                                            Page 2 of 2

The specific shorted voltage buffer card provided the signal to the NNI "x"
saturation meter. 

Licensees which utilize Bailey Control Company Voltage Buffer Cards are
requested to carefully inspect the cards for possible misalignment and take
corrective actions if misalignments are identified. Specific instructions 
for carrying out these inspections and providing any other information which 
may be required to define appropriate corrective action is being prepared by
Baily Control Company for transmittal to purchasers of this equipment by 
March 11, 1980. 

Initial screening of IE Bulletin No. 79-27 responses indicates a range of
responses regarding depth and scope of review.

IE Bulletin No. 79-27 was intended to cause licensees to investigate loss of 
individual power supplies as well as total loss of an inverter or vital bus.
An addendum to IE Bulletin No. 79-27 is planned to be issued in the near
future to reflect the CR-3 event.

This Information Notice is provided to inform licensees of a possibly
significant matter. It is expected that recipients will disseminate the
information to all operational personnel working at their licensed
facilities. (A meeting was held with B&W licensees in NRC Headquarters on
March 5, 1980 to review the event at Crystal River and to discuss proposed
corrective actions. Responses to specific questions have been requested of
the-B&W licensees.) If you have questions regarding this matter, please
contact the Director of the appropriate NRC Regional Office.

No written response to this Information Notice is required.

Enclosure: 
Sequence of Events
.

                 26 FEBRUARY TRANSIENT CRYSTAL RIVER UNIT-3 
                             SEQUENCE OF EVENTS 

EVENT SYNOPSIS 

At 14:23 on February 26, 1980, Crystal River-3 Nuclear Station experienced
a reactor trip. Nominal full power primary and secondary system parameters
were present. A synopsis of key events and parameters was obtained from the
plant computer's post-trip review and plant alarm summary, the sequence of
events, monitor, control room strip charts, and the Shift Supervisor's log.

The reactor was operating at approximately 100% full power with Integrated
Control System (ICS) in automatic. No tests were in progress.

     Time           Event                         Cause/Comments 

14:23:21       +24 Volt Bus Failure (NNI power    The positive 24 VDC bus
               loss "X" supply)                   shorted dragging the bus
                                                  voltage down l to a low
                                                  voltage trip condition. 
                                                  There is a built-in 1/4 to
                                                  1/2 second delay at which
                                                  time all power supplies
                                                  will trip. The trip
                                                  indication on negative (-) 
                                                  voltage was missed by the 
                                                  annunciator. Following the
                                                  NNI power failure, much of
                                                  the control room 
                                                  indication was lost. Of the 
                                                  instrumentation that
                                                  remained operable,
                                                  transient conditions made
                                                  their indications 
                                                  questionable to the
                                                  operators.

14:23:21       PORV and Pressurizer Spray         When the positive 24 VDC
               Valve Open                         supply was lost due to the 
                                                  sequence discussed above
                                                  the signal monitors in NNI
                                                  changed state causing
                                                  PORV/Spray valves to open.
                                                  The PORV circuitry is 
                                                  designed to seal in upon
                                                  actuation and did so. The
                                                  resultant loss of the
                                                  negative 24 VDC halted
                                                  spray valve motor operator,
                                                  and prevented PORV seal-in
                                                  from clearing on low 
                                                  pressure. It is postulated 
                                                  that the PORV opened fully
                                                  and the spray valve stroked
                                                  for approximately 112
                                                  second. The "40% open"
                                                  indication on spray valve
                                                  did not actuate, therefore,
                                                  the spray valve did not
                                                  exceed 40% open.
.

26 February Transient CR-3           -2- 

     Time           Event                         Cause/Comments 

14:23:21       Reduction in Feedwater             As a result of the "X"
                                                  power supply failure many
                                                  primary plant control
                                                  signals responded
                                                  erroneously. T-cold failed
                                                  to 570F (normal
                                                  indication was  557F)
                                                  producing several spurious
                                                  alarms. T-ave failed to
                                                  570F (decreased). The
                                                  resultant T-ave error
                                                  modified the reactor demand
                                                  such that control rods were
                                                  withdrawn to increase T-ave
                                                  and reactor power. The
                                                  power increase was
                                                  terminated at 103% by the
                                                  ICS and a "Reactor Demand
                                                  High Limit" alarm was
                                                  received T-hot failed to
                                                  570 F (low) and RC flow
                                                  failed to 40 x 106 lbs/hr
                                                  in each loop (low). Both
                                                  these failures created a
                                                  BTU alarm and limit on
                                                  feedwater which reduced
                                                  feedwater flow to both
                                                  OTSG's to essentially zero.
                                                  Turbine header pressure
                                                  failed to 900 psig (high)
                                                  which caused the turbine
                                                  valves to open slightly to
                                                  regulate header pressure
                                                  thus increasing generated
                                                  megawatts. These combined
                                                  failures resulted in a loss
                                                  of heat sink to the reactor
                                                  initiating high RCS
                                                  pressure condition.

14:23:35       Reactor Trip/Turbine Trip          Rx trip caused by high RCS
                                                  pressure at 2300 psi.
                                                  Turbine was tripped by the
                                                  reactor. 

14:24:02       Hi Pressure Inj.                   This was a computer
               Reg. (Flag)                        printout and indicates <
                                                  50 subcooling.  (The
                                                  lowest level of subcooling
                                                  was 8F for a very
                                                  short period of time, at
                                                  about 14:30) 

14:24:02       Loss of Both Condensate            Suspect that the condensate
               Pumps                              pump tripped due to high
                                                  De-aeriating feed tank
                                                  level.
.

26 February Transient CR-3          - 3 -

     Time           Event                         Cause/Comments 

14:24:02                                          This is verified by a 
(Continued)                                       series of questions marks 
                                                  ???? printed by computer 
                                                  indicating that the level 
                                                  instrument was over-ranged.
                                                  A low flow indication in
                                                  the gland steam condenser
                                                  was also indicated by
                                                  computer.

14:25:50       PORV Isolated                      At this time a high RC
                                                  Drain Tank level alarm was 
                                                  received. This was
                                                  resultant from the PORV 
                                                  remaining open and was 
                                                  positive indication that 
                                                  the PORV was open. At this 
                                                  time, the operator closed 
                                                  the PORV block valve due
                                                  to RCS pressure decreasing
                                                  and high RCDT level. 

14:26:41       HPI Auto Initiation                HPI initiated automatically 
                                                  due to low RCS pressure of 
                                                  1500 psig. The low 
                                                  pressure condition was 
                                                  resultant from the PORV 
                                                  remaining full open while 
                                                  the plant was tripped. 
                                                  Full HPI was initiated with
                                                  3 pumps resulting in 
                                                  approximately 1100 gpm flow 
                                                  to the RCS. At this time, 
                                                  all remaining non-essential 
                                                  reactor building (RB) 
                                                  isolation valves were
                                                  closed per TMI Lessons 
                                                  Learned Guidelines.

14:26:54       RC Pumps Shutdown                  Operator turned RC pumps 
                                                  off as required by the 
                                                  applicable emergency 
                                                  procedure and B&W small 
                                                  break guidelines.

14:27:20       RB Pressure Increasing             This is first indication 
                                                  that RCDT rupture disc had 
                                                  ruptured. RB pressure 
                                                  increase data was obtained 
                                                  from Post Trip Review and 
                                                  Strip Chart indication.
.

26 February Transient CR-3          - 4 - 

     Time           Event                         Cause/Comments 

14:31:32       RB Pressure High                   This alarm was initiated
                                                  by 2 psig in RB. This is 
                                                  attributed to steam release 
                                                  from RCDT. Code safeties
                                                  had not opened at this time
                                                  based upon tail pipe
                                                  temperatures recorded at
                                                  14:32:03 (Computer).

14:31:49       OSTG "A" Rupture Matrix            This occurred due to < 600
               Actuation                          psig in OTSG "A". The low
                                                  pressure was caused by OTSG
                                                  "A" boiling dry which was
                                                  resultant from the BTU
                                                  limit and failed power 
                                                  supply to OTSG "A" level 
                                                  transmitter. This resulted
                                                  in the closure of all
                                                  feedwater and Steam block
                                                  valves which service OTSG
                                                  "A".

14:31:59       Main Feedwater Pump 1A             Caused by suction valve
               Tripped                            shutting due to rupture
                                                  matrix actuation in
                                                  previous step.

14:32:14:41    ES A/B Bypass                      Manually bypassed and HPI 
                                                  balanced between all 4
                                                  nozzles (Total flow
                                                  approximately 1100
                                                  gpm-small break operating 
                                                  guidelines).

14:32:35       Started Steam Driven               Started by operator to
               Emergency Feedwater Pump           ensure feedwater was
                                                  available to feed OTSG's.

14:33          Core Exit Temp. Verified           The core exit incore 
                                                  thermocouples indicated
                                                  the highest core outlet
                                                  temperature value was
                                                  560F. RCS pressure 
                                                  was 2353 psig at this time, 
                                                  therefore, the subcooling 
                                                  margin at this time was
                                                  100F. Minimum
                                                  subcooling margin for  the
                                                  entire transient was
                                                  8F at 14:30. It is
                                                  postulated that some
                                                  localized boiling occurred
                                                  in the core at this point
                                                  as indicated by the self 
                                                  powered neutron detectors.
.

26 February Transient CR-3          - 5 - 

     Time           Event                         Cause/Comments 

14:33:14:44    Started Motor Driven Emer-         Same discussion as "Started
               gency Feedwater Pump               Steam Driven Emergency 
                                                  Feedwater Pump." 

14:33:30       RC Pressure High (2395             At this point, pressurizer
               psig)                              is solid and code safety
                                                  lifts (RCV-8). This is the
                                                  highest RCS pressure as
                                                  recorded on Post Trip
                                                  Review.  Apparently, RCV-8
                                                  lifted early due to seat
                                                  leakage prior to the
                                                  transient and RCV-9 did not
                                                  lift.

14:34:23       RB Dome Hi Rad Level               RMG-19 alarmed at this 
                                                  point. Highest level 
                                                  indicated during course of 
                                                  incident was 50 R/hr. High 
                                                  radiation levels in RB
                                                  caused by release of non-
                                                  condensable gases in the
                                                  pressurizer and coolant.

14:35:13       Attempted NNI Repower              This resulted in spikes
               Without Success                    observed on de-energized
                                                  strip charts.

14:36:50       Computer Overload                  Caused by overload of 
                                                  buffer. Resulted 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 
                                                  Operating Range.

14:44:12       NNI Power Restored                 NNI was restored by
               Successfully                       removing the "X"-NNI Power
                                                  Supply Monitor Module. This
                                                  allowed the breakers to be
                                                  reclosed. At this time, it
                                                  was observed that the "A"
                                                  OTSG was dry, the
                                                  pressurizer was solid
                                                  (Indicated off-scale high), 
                                                  RC outlet temperature 
                                                  indicated 556F (loop
                                                  A & B  average), and RC
                                                  average temperature
                                                  indicated 532F  (Loop
                                                  A & B). The highest core
                                                  exit thermocouple
                                                  temperature at
.

26 February Transient CR-3          - 6 - 

     Time           Event                         Cause/Comments

14:44:12                                          this time was 531F.
(Continued)                                       RCS pressure was 2400 psig 
                                                  (saturation temp. at this 
                                                  pressure is 662F.).
                                                  This data verified natural
                                                  circulation was in Progress
                                                  and the plant subcooling
                                                  margin was 131F 
                                                  (based on core exit 
                                                  thermocouples).

14:44:31       RB Isolation and Cooling           At this time, RB pressure
               Actuation                          increased to 4 psig and 
                                                  initiated RB Isolation. The
                                                  operator verified all
                                                  immediate actions occurred
                                                  properly for HPI, LPI, and
                                                  RB Isolation and Cooling.
                                                  The increasing RB pressure
                                                  was resultant from RCV-8
                                                  relieving pressure due to 
                                                  continued HPI.

14:46:10       Bypassed HPI, LPI and RB           These "ES" systems were
               Isolation and Cooling              bypassed at this time to 
                                                  balance HPI flow and
                                                  restore cooling water to
                                                  essential auxiliary
                                                  equipment (i.e., RCP's,
                                                  letdown coolers, CRDM's 
                                                  etc.).

14:51:57       Rupture Matrix Actuation           The actuation was resultant
               on OTSG-B                          from a degradation of
                                                  OTSG-B pressure. Cold
                                                  emergency feed was being
                                                  injected into the OTSG at
                                                  this time. This matrix
                                                  actuation isolated all
                                                  feedwater and steam block 
                                                  valves to the B-OTSG and
                                                  tripped the "B" main FW
                                                  pump. Both Emergency FW
                                                  pumps were already in
                                                  operation at this time.
                                                  B-OTSG level at this time
                                                  was 70% (Operation Range).

14:52          HPI Throttled and RCS              At this time, the maximum
               Pressure Reduced to 2300           core exit thermocouple 
               psig                               temperature was 515F,
                                                  RCS  pressure was 2390
                                                  psig.
.

26 February Transient CR-3          - 7 - 

     Time           Event                         Cause/Comments 

14:52                                             Therefore, the subcooling
(Continued)                                       margin was 147F.
                                                  Natural circulation was in
                                                  effect as verified
                                                  previously. All conditions
                                                  had been satisfied to
                                                  throttle HPI. Therefore, 
                                                  flow was throttled to 
                                                  approximately 250 gpm to 
                                                  reduce RCS pressure to 2300 
                                                  psig in order to attempt
                                                  to  reduce the flow rate
                                                  through  RCV-8 and into the
                                                  RB.

14:53          Reestablished Letdown              At this time, the operator 
                                                  was attempting to establish 
                                                  RCS pressure control via 
                                                  normal RC makeup and
                                                  letdown.

14:56          Opened MU Pump Recirc.             This was done to assure the
               Valves                             MU pumps would have minimum 
                                                  flow at all times to
                                                  prevent  possible pump
                                                  damage.

14:56:43       Bypassed the A-OTSG                Feedwater was slowly
               Rupture Matrix and                 admitted to the A-OTSG
               Reestablished Feed                 which was dry up to this
               to the A-OTSG                      point. Feedwater was
                                                  admitted through the
                                                  Auxiliary FW header via
                                                  the EFW bypass valves. The
                                                  feedrate was very slow in
                                                  order to minimize thermal
                                                  shock to the OTSG and
                                                  resultant depressurization
                                                  of the RCS. RCS pressure
                                                  control was very unstable
                                                  at this time.

14:57:09       Bypassed the B-OTSG                This was done to regain FW
               Rupture Matrix                     control of the B-OTSG.
                                                  Level was still high in
                                                  this OTSG (approximately
                                                  65% Operating Range).
                                                  Therefore, feed was not
                                                  necessary at this time. The
                                                  Main Steam Isolation valves
                                                  were open in preparation
                                                  for bypass valve operation
                                                  (when necessary).
.

26 February Transient CR-3          - 8 - 

     Time           Event                         Cause/Comments 

14:57:15       Established RC Pump                This was done in
               Seal Return                        preparation for a RCP start
                                                  (when necessary)  and to
                                                  minimize pump seal 
                                                  degradation.

15:00-09       Reestablished Level                This verified feedwater was
               in A-OTSG                          being admitted to the OTSG 
                                                  and made it available for
                                                  core cooling via natural 
                                                  circulation. Feed to this 
                                                  generator was continued
                                                  with the intent of
                                                  proceeding to 95% on the
                                                  Operating Range.

15:00-09       77F Subcooled "A" Loop        This value was based upon
                                                  "A"  RCS loop parameters
                                                  at this  time. The "A" loop
                                                  was being  cooled down at
                                                  this time by the  A-OTSG
                                                  fill and the operator was 
                                                  attempting to equalize loop
                                                  temperatures.

15:15          23F Delta-T/Manned the        At this time, loop
               Technical Support Center           temperatures were nearing
                                                  equalization. This  delta-T
                                                  was calculated from loop 
                                                  A & B T C's and core exit 
                                                  thermocouples.

15:17          Declared Class "B"                 This was done based on the
               Emergency                          fact there was a loss of
                                                  coolant through RCV-8 into
                                                  the containment and HPI
                                                  had been initiated. All
                                                  non-essential CR-3
                                                  personnel were directed to 
                                                  evacuate and contact of
                                                  off-site agencies began.
                                                  Survey team was sent to
                                                  Auxiliary Building.

15:19          Opened Emergency FW                At this point the A-OTSG
               Block to B-OTSG                    level was increasing and
                                                  the decision was made to
                                                  commence filling the B-OTSG 
                                                  simultaneously. The intent
                                                  was to go 95% on both 
                                                  OTSG's without exceeding
                                                  RCS cooldown limits (100
                                                  F/hr) while maintaining
                                                  RCS pressure control.
.

26 February Transient CR-3          - 9 - 

     Time           Event                         Cause/Comments 

15:26          Lo Level Alarm in Sodium           This was resultant from the
               Hydroxide Tank                     tank supply valve opening
                                                  when the 4 psig RB
                                                  isolation and cooling
                                                  signal actuated. The 
                                                  sodium hydroxide was
                                                  released to both LPI
                                                  trains. Sodium Hydroxide
                                                  was admitted to the RCS
                                                  via HPI from the BWST. 
                                                  (Approximately 2 ppm
                                                  injected into the RCS.) 

15:50          Terminated HPI                     At this time, all
                                                  conditions had been
                                                  satisfied (per small break
                                                  operating guidelines) to
                                                  terminate HPI. RCS 
                                                  pressure control had been 
                                                  established using normal 
                                                  makeup and letdown. HPI
                                                  was terminated and
                                                  essentially all releases
                                                  to the RB were
                                                  discontinued.

16:00          Commenced Pressurizer              At this time, RCS pressure
               Heatup                             and temperature were well
                                                  under control. Natural
                                                  circulation was functioning
                                                  as-designed (approximately
                                                  23F delta-T).  RCS
                                                  temperature was being 
                                                  maintained at approximately 
                                                  450F. RCS pressure was 
                                                  approximately 2300 psig.
                                                  The decision was made at
                                                  this point to commence
                                                  pressurizer heatup in
                                                  preparation to re-establish
                                                  a steam space in the
                                                  pressurizer. 

16:07          Survey Team Report                 The Emergency Survey Team
                                                  reported no radiation
                                                  survey  results taken
                                                  offsite were above
                                                  background.

16:08:04       Shutdown Steam Driven              The motor driven Emergency
               Emergency FW Pump                  FW pump was running,
                                                  therefore, the steam driven
                                                  pump was not needed. The
                                                  plant remained in this
                                                  condition for approximately
                                                  2 hours, while heating up
                                                  the pressurizer to
                                                  saturation temperature for 
                                                  1800 psig.
.

26 February Transient CR-3          -10- 

     Time           Event                         Cause/Comments 

18:05          Established Steam Space            At this point, pressurizer
               in Pressurizer                     temperature was
                                                  approximately 620F. 
                                                  Pressurizer level was 
                                                  brought back on scale by 
                                                  increasing letdown. From 
                                                  this point pressurizer 
                                                  level was reduced to 
                                                  normal operating level and 
                                                  normal pressure was 
                                                  established via 
                                                  pressurizer heaters.

18:30          Terminated Class B                 State and Federal Agencies
                Emergency                         notified.

21:07          Forced Flow Initiated              The decision was made to
               in RCS                             re-establish forced flow 
                                                  cooling in the RCS at this 
                                                  time. B&W and NRC were 
                                                  consulted. RCP-1B and 1D 
                                                  were started. At this 
                                                  point, RCS parameters were 
                                                  stabilized and maintained 
                                                  at RC pressure-2000 psig,
                                                  RCS temperature-420F.
                                                  Pressurizer level-235 
                                                  inches. The plant was
                                                  considered in a normal 
                                                  configuration. 
.
```

*Page Last Reviewed/Updated Friday, May 22, 2015*

#### Home

            - News Releases
            - Event Reports
            - ADAMS
            - Open Gov
            - Digital Government
            - The Student Corner
            - Photos & Video
            - For Developers

#### About Us

            - Strategic Plan
            - Budget & Performance
            - Agency Financial Report
            - License Fees
            - History of the NRC
            - Career Opportunities
            - NRC Ethics
            - Agency Status
            - Contact Us

#### Popular Documents

            - Info Digest
            - Fact Sheets & Brochures
            - Forms
            - Electronic Submittals Application
            - Adjudicatory Submissions
            - NRC Reports – NUREG
            - NRC Regulations – 10-CFR
            - Inspection Reports
            - Plain Writing
            - Enforcement Actions
            - RULEMAKING

#### Stay Connected

            - Blog
            - Facebook
            - Twitter
            - YouTube
            - Flickr
            - GovDelivery
            - RSS

Regulations.govUSA.govRecoveryFOIANo Fear EEOInspector General ProgramSite MapAccessibilityPrivacy PolicySite DisclaimerFor Employees