United States Nuclear Regulatory Commission - Protecting People and the Environment

ACCESSION #: 9407070001
                       LICENSEE EVENT REPORT (LER)

FACILITY NAME:  BYRON NUCLEAR POWER STATION               PAGE: 1 OF 07

DOCKET NUMBER:  05000454

TITLE:  OPERABILITY DETERMINATION OF SOURCE RANGE NUCLEAR
        INSTRUMENTATION

EVENT DATE: 07/01/92    LER #: 92-004-01    REPORT DATE: 06/28/94

OTHER FACILITIES INVOLVED:  Byron, Unit 2           DOCKET NO:  05000455

OPERATING MODE:  1   POWER LEVEL:  100

THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR
SECTION:
50.73(a)(2)(v)

LICENSEE CONTACT FOR THIS LER:
NAME:  K. ELAM,                             TELEPHONE:  (815) 234-544
       LEAD NUCLEAR ENGINEER, EXT. 2247

COMPONENT FAILURE DESCRIPTION:
CAUSE:  B   SYSTEM:       COMPONENT:       MANUFACTURER:
REPORTABLE NPRDS:  N

SUPPLEMENTAL REPORT EXPECTED:  NO

ABSTRACT:

At 1505 on 07/01/92, Byron Station received an operability assessment,
ENC-QE-40.1, regarding the Boron Dilution Protection System (BDPS) (NR)
[IG].  The operability assessment was precipitated by the discovery of
two non-conservative assumptions in the safety analysis for the system.
On-Site Review 92-089 was immediately convened that concurred with the
determination that BDPS is to be considered operable under a certain set
of conditions.  However, when the plant is outside of these conditions,
the BDPS subsystem may not be capable of performing its intended safety
function.  Special Operating Order SO-U1/U2-19 was revised to implement
the findings and recommendations of the operability assessment by
detailing the conditions necessary for BDPS operability.  This Special
Operating Order remained in effect until further safety analysis was
performed that provided permanent resolution of this issue.

This event is reportable pursuant to 10CFR50.73(a)(2)(v), any event or
condition that alone could have prevented the fulfillment of the safety
function to structures or systems that are needed to mitigate the
consequences of an accident.

     (9941R\WPF\061694-2)

END OF ABSTRACT

TEXT                                                          PAGE 2 OF 7

A.   PLANT CONDITIONS PRIOR TO EVENT:

     Event Date/Time 07/01/92 / 1505

     Unit 1 MODE 1 - Operations         Rx Power   100%

               RCS [AB] Temperature/Pressure 580 degrees F/2235 psig.

     Unit 2 MODE 1 - Operations         Rx Power   100%

               RCS [AB] Temperature/Pressure 579 degrees F/2238 psig.

B.   DESCRIPTION OF EVENT:

     At 1505 on 07/01/92, Byron Station received an operability
     assessment, ENC-QE-40.1, regarding the Boron Dilution Protection
     System (BDPS) (NR) [IG].  On-Site Review 92-089 was immediately
     convened that concurred with the determination that BDPS is to be
     considered operable under a certain set of conditions.  However,
     when the plant is outside of these conditions, the BDPS subsystem
     may not be capable of performing its intended safety function.
     Special Operating Order SO-U1/U2-19 was revised to implement the
     findings and recommendations of the operability assessment by
     detailing the conditions necessary for BDPS operability.  This
     Special Operating Order remained in effect until further safety
     analysis was performed that provided permanent resolution of this
     issue.

     On March 4, 1992, Westinghouse issued a Potential Issue (PI) on the
     operability of the Boron Dilution Protection System.  This PI was
     issued because two potential non-conservatism were identified in the
     original Safety Analysis for this system:

     1.   The assumed Inverse Contrite Ratio (ICRR) curve in the analysis
          was found to be non-conservative at another Westinghouse plant.

     2.   The setpoint for the flux doubling did not include an
          uncertainty analysis.

     At the time the PI was received from Westinghouse, insufficient
     information was available to determine operability of the system.
     Pursuant to the Pl issued by Westinghouse, Byron Station, in concert
     with Nuclear Fuel Services (NFS), Nuclear Licensing (NLA), and
     Braidwood Station agreed on the conservative compensatory actions
     included in OSR 92-032.  These actions mitigated the probability and
     consequences of a dilution accident by maintaining a high shutdown
     margin and administratively controlling the valves capable of
     contributing to an inadvertent dilution.  These actions were:

     Whenever either unit was in Modes 3, 4, or 5:

     1.   The required shutdown margin was increased to a minimum of
          1.65% (from 1.0%) when in Mode 5.

     (9941R\WPF\061694-3)

TEXT                                                          PAGE 3 OF 7

B.   DESCRIPTION OF EVENT: (continued)

     2.   Normal shutdown operating practice was to maintain charging
          flow less than 130 gpm.  If charging flow was to be maintained
          at greater than 130 gpm, the shutdown margin was increased to:

               Mode 4: 1.45%
               Mode 5: 1.84%

     3.   Manual valve _BR7004 to the primary water system was locked
          closed.

     4.   Administrative controls were implemented that required the
          possible dilution paths be isolated (valves_CV8428, _CV8435,
          _CV8441, _CV8439 locked closed and verified closed and air or
          electrical power remove from _CV111B) before draining the
          pressurizer level below the bottom of the indicated range while
          in Mode 5.

     5.   Administrative controls were implemented that required the
          Boron Thermal Regeneration System (BTRS) be isolated prior to
          draining the pressurizer level below the bottom of the
          indicated range while in Mode 5, and that the demineralized
          water supply valve for the demineralized flush be locked
          closed.  Also that demineralized flush operations performed
          while in Mode 5 only be performed under strict administrative
          procedure, such additional valves be closed and written
          verification and independent check be obtained that the valves
          to the primary water system or demineralized water supply were
          result and locked after flushing operations.
          (_BR7052, _BR7053, _BR7054, _CV8542)

     6.   Flushing the emergency boration line with primary water was
          strictly controlled and only when the charging rate was
          monitored and controlled to less than 130 gpm.

     7.   The outlet valves from the Boric Acid Storage Tanks were
          verified open after any maintenance activities. (_AB8461)

Since that time, Nuclear Fuel Services (NFS) and Engineering and Nuclear
Construction (ENC) have pursued evaluating the operability of the system,
and concluded that the generic concerns for the BDPS system are
applicable to Byron:

     1.   The assumed ICRR curve does not bound the Byron and Braidwood
          sites.  It was found that the curve from Braidwood Unit 1 Cycle
          3 has been the most bounding thus far, and that it will likely
          remain bounding".

     2.   A sensitivity analysis had not been performed for the Byron and
          Braidwood sites.  Although it has not been possible to provide
          a quantitative uncertainty for the circuitry at this time, a
          best estimate of the uncertain for the doubling setpoint is
          30%, thus making the analysis setpoint 2.6.

     (9941R\WPF\061694-4)

TEXT                                                          PAGE 4 OF 7

B.   DESCRIPTION OF EVENT: (continued}

     Through the performance of specialized safety analysis cases,
     Nuclear Fuel Services (NFS) concluded that BDFS remains OPERABLE in
     certain conditions.  However, the analysis failed to demonstrate
     operability for all conditions.  If all of these conditions are not
     met, the system is to be considered INOPERABLE.  The condition are:

     1.   The Shutdown Margin must be at least 1300 pcm in Modes 3, 4,
               and 5.

     2.   All Loop Stop Isolation Valves must be open.

     3.   At least 1 Reactor Coolant Pump must be operating.

     4.   The Source Range Nuclear Instrumentation Count Rate must be at
          least 10 counts per second.

     With the preceding conditions not being met, both trains of BDPS
     shall be declared inoperable and the appropriate Technical
     Specification actions taken.

     This issue is reportable under Title 10, Code of Federal
     Regulations, Part 50, Section 73, (a)(2)(v), any event a condition
     that alone could have prevented the fulfillment of the safety
     function of structures or systems that a needed to mitigate the
     consequences of an accident.

C.   CAUSE OF EVENT:

     The cause for this event was inadequate safety analysis and
     subsequent review for the Boron Dilution Protection System.

     At the time of the original analysis, Westinghouse used the most
     limiting ICRR available from the industry in the input assumptions
     to the.postulated accidents.  However, development of new low
     leakage loading patterns and neutron source positions have rendered
     that ICRR non-bounding.

     It is not known exactly why an instrument uncertainty analysis was
     not included in the design of the BDPS setpoint.  However, it is
     believed that the fact that BDPS was not a part of the original
     design of the plant and that BDPS does not have its own Limiting
     Condition for Operability in the Byron/Braidwood Technical
     Specifications contributed to this oversight.

     (9941R\WPF\061694-5)

TEXT                                                          PAGE 5 OF 7

D.   SAFETY ANALYSIS:

     It has been concluded that BDPS may be incapable of performing its
     intended safety function in the event of a boron dilution accident
     under certain plant conditions.  However, the safety analysis
     performed merely failed to demonstrate acceptable performance for
     all conditions using the present analysis method.  After
     implementing possible improvements to the method of analysis, a
     wider spectrum of conditions may be acceptable for BDPS operability.

     Had certain plant conditions existed where the BDPS system was
     inoperable and a dilution accident was initiated two other sources
     for indication of the decrease in shutdown margin were available to
     alert the operator.  During shutdown conditions, the Source Range
     indication is broadcast audibly in the control room and containment.
     Also, the High Flux at shutdown annunciator, which is intended to
     notify personnel of an inadvertent criticality during fuel load and
     is set to actuate at an instantaneous indication of 5 times the
     background contrite, is available in Modes 3 through 6.

     Furthermore, the consequences of an unmitigated dilution accident do
     not pose a substantial safety hazard.  Analysis performed by Los
     Alamos National Laboratory (LANL) for the NRC has concluded that an
     unmitigated dilution of a PWR in a shutdown Mode would result in a
     return to power and may result in an increase in reactor coolant
     system pressure and some fuel damage.  LANL further concluded that
     the return to power transient we be self limiting by virtue of the
     inherent negative feedback of the reactor.  The self limiting return
     to power would also limit fuel damage and repressurization.

E.   CORRECTIVE ACTIONS:

     Upon the notification of this concern to Byron, the compensatory
     actions documented under OSR 92-032 were promptly implemented.

     Upon the receipt of the Operability Assessment from NFS specifying
     the conditions necessary for BDPS operability, Byron Station
     immediately implemented the following actions:

     1.   The Special Operating Order (SO Unit 1 /Unit 2 92-019) was
          revised to implement the four conditions for operability.

     2.   The station's Nuclear Regulatory Commission Resident Inspector
          was notified of this condition.

     3.   The station made the required Emergency Notification System
          phone call within the required 4 hours.

     (9941R\WPF\061694-6)

TEXT                                                          PAGE 6 OF 7

E.   CORRECTIVE ACTIONS: (continued)

     Future actions were necessary to resolve this issue for the
     long-term.  A synopsis of these actions included the following:

     1.   The 30% uncertainty of the doubling setpoint was confirmed by
          ENC.  Westinghouse performed an analysis that supported the 30%
          conclusion and the NRC accepted that value when approving the
          new BDPS Tech Spec. (NTS # 454-200-92-03200-01)

     2.   Byron Station determined the maximum primary flow rate through
          flow orifice _CV17M by measurement.  The highest measured flow
          rate was 92 gpm on Unit 1.  The analysis assumed 150 gpm,
          therefore, future analysis may benefit from using a lower flow
          rate.  (NTS # 454-200-92-03200-02)

     3.   NFS investigated the possibility of demonstrating a wider array
          of operable conditions for BDPS.  A partners was formed with
          Comanche Peak, Wolf Creek, Callaway, Braidwood and Byron, with
          Comanche Peak being the lead plant in any licensing changes.
          Comanche Peak was successful in licensing the Volume Control
          Tank high level alarm to render the Boron Dilution Protection
          System moot in analysis space.  Since Byron and Braidwood had
          already obtained approval on the new BDPS Tech Spec, Byron will
          follow Comanche Peak's lead should that avenue pay economic
          dividends. (NTS # 454-200-92-03200-03)

     4.   Amendment 51 to the Byron and Braidwood Technical
          Specifications was approved by the NRC that includes new LCO
          for BDPS.  Tech Spec 3/4.1.2.7 reflects the conditions for
          operability and includes appropriate action statements and
          surveillance requirements. (NTS # 454-200-92-03200-04)

F.   RECURRING EVENTS SEARCH AND ANALYSIS:

     a)   EVENT SEARCH (DIR, LER)

          There have been no previous occurrences of a DVR caused by
          improper analysis of the BDPS system, although non-conservatism
          in the analysis of this system have occurred in the past.

     b)   INDUSTRY SEARCH (OPEX's NPRDS)

          NPRDS is not applicable for this event, however, this event was
          initiated by a discovery at Comanche Peak.  The other stations
          directly affected by the BDPS analysis are Braidwood, Callaway,
          and Wolf Creek.

          OPEX: Plant Status Report (PS)#2607.

     (9941R\WPF\061694-7)

TEXT                                                          PAGE 7 OF 7

F.   RECURRING EVENTS SEARCH AND ANALYSIS: (continued)

     c)   NWR

          Not applicable.

     d)   ANALYSIS

          No trend identified.

G.   COMPONENT FAILURE DATA:

                                             MODEL     MFG PART
     MANUFACTURER        NOMENCLATURE        NUMBER    NUMBER

     No equipment failed during this event.

H.   OTHER RELATED DOCUMENTS:

     ENC-QE-40.1, Operability Determination Checklist
     OSR 92-032, Precautionary Measures Taken for BDPS Potential Issue
     OSR 92-089, Review of ENC-QE-40.1 for BDPS Operability

I.   EFFECTIVENESS REVIEW:

     Not applicable.

J.   ADDITIONAL DATA:

     a)   Affected Technical Specification: 3/4.3.1.1, Functional Unit 6

     b)   Procedures: Not applicable

     c)   Cause Code: BD2.6

     d)   Equipment Involved: Boron Dilution Protection System of the
          Source Range Instrumentation

     e)   Other: BDPS, Source Range, Westinghouse Analysis

     (9941R\WPF\061694-8)

ATTACHMENT TO 9407070001                                      PAGE 1 OF 2

     Commonwealth Edison
     Byron Nuclear Station
     4450 North German Church Road
     Byron, Illinois 61010

June 28, 1994

LTR:      BYRON 94-0238
FILE:     3.03.0800 (1.10.0101)

U.S. Nuclear Regulatory Commission
Document Control Desk
Washington, D.C. 20555

Dear Sir:

     The Enclosed Supplemental Licensee Event Report from Byron
Generating Station is being transmitted to you in accordance with the
requirements of 10CFR50.73(a)(2)(v).

     This report is number 92-004; Docket No. 50-454.

Sincerely,

G. K. Schwartz
Station Manager
Byron Nuclear Power Station

GKS/DSK/ng

Enclosure:     Licensee Event Report No. 92-004 Supplement

cc:  J. Martin, NRC Region III Administrator
     NRC Senior Resident Inspector
     INPO Record Center
     CECo Distribution List

     (4347z/VS/100593)

ATTACHMENT TO 9407070001                                      PAGE 2 OF 2

                 SIGNATURE PAGE FOR LICENSE EVENT REPORT

                               LER Number
                               454: 92-004

Title of Event:     Operability Determination of Source Range Nuclear
                    Instrumentation

Occurred:      07-01-92/ 1505
                 Date    Time

OSR DISCIPLINES REQUIRED:
                                                       SES       DATE

Acceptance by Station Review:

          OE             Date           SES            Date

          RAS            Date           OTHER          Date

Approved by:

          Station Manager          Date

     (9941R\WPF\061694-1)

*** END OF DOCUMENT ***


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