- About NRC           - The Commission
      - Statutory Authority
      - Strategic Plan
      - Budget & Performance
      - Organization & Functions
      - Locations
      - History
      - Values
      - Policymaking
      - Radiation Protection
      - Fire Protection
      - Regulation
      - Emergency Preparedness   and Response
      - Public Affairs
      - Congressional Affairs
      - International Programs
      - State and Tribal Programs
      - Alternative Dispute   Resolution Programs
      - Contact Us
      - Employment
      - Contracting
      - Grants
      - Civil Rights
      -     
    - NuclearReactors           - Power Reactors
      - Research and Test Reactors
      - Operating Reactors
      - Operator Licensing
      - New Reactors
      - Advanced Reactors
      - Operator Licensing for New Reactors
      - Specific Plants of Interest
      -     
    - NuclearMaterials           - Special Nuclear Material
      - Source Material
      - Byproduct Material
      - Med, Academic & Ind Uses
      - Source Materials Facilities
      - Uranium Recovery Facilites
      - Fuel Cycle Facilities
      - Materials Transportation
      -     
    - RadioactiveWaste           - Low-Level Waste
      - Waste Incidental to Reprocessing
      - High-Level Waste
      - Uranium Mill Tailings
      - LLW Disposal
      - HLW Disposal
      - Storage of Spent Nuclear Fuel
      - Transporation of Spent Nuclear Fuel
      -     
    - NuclearSecurity           - Domestic Safeguards
      - Information Security
      - Radioactive Material
      - &;   
    - Public Meetings& Involvement           - Open Government
      - Public Meeting Schedule
      - Commission Schedule
      - Conferences & Symposia
      - Documents for Comment
      - Involving Stakeholders
      - Info Quality Guidelines
      - Subscribe to E-mail Notices 
      -     

```
                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                     OFFICE OF NUCLEAR REACTOR REGULATION
                         WASHINGTON, D.C.  20555-0001

                               October 30, 1997


NRC INFORMATION NOTICE 97-76:  DEGRADED THROTTLE VALVES IN EMERGENCY
                               CORE COOLING SYSTEM RESULTING FROM
                               CAVITATION-INDUCED EROSION DURING A
                               LOSS-OF-COOLANT ACCIDENT


Addressees  

All holders of operating licenses for pressurized-water reactors except those
who have permanently ceased operations and have certified that fuel has been
permanently removed from the reactor vessel.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to potential problems caused by degradation of
emergency core cooling system (ECCS) throttle valves in the intermediate-head
safety injection pump hot-leg and cold-leg flow paths and in the charging pump
(high-head safety injection) cold-leg flow paths during certain loss-of-
coolant-accident (LOCA) scenarios.  It is expected that recipients will review
the information for applicability to their facilities and consider actions, as
appropriate, to avoid similar problems.  However, suggestions contained in
this information notice are not NRC requirements; therefore, no specific
action or written response is required.  

Description of Circumstances

On January 11, 1996, Westinghouse issued Nuclear Safety Advisory Letter (NSAL)
96-001, "Erosion of Globe Valves in ECCS Throttling Applications," which
alerted Westinghouse plant owners to the potential cavitation and erosion of
throttling valves used in high- and intermediate-head ECCS.  The valve erosion
occurs as a result of high differential pressure and flow that may occur
during a LOCA.  As a result of the valve erosion, the high- and intermediate-
head safety injection pumps may exceed their run out limits, thus forcing the
operator to secure the pumps before the time established in the licensing
basis for pump operability.  This issue was originally identified at Sequoyah
Nuclear Plant through the corrective action program generic review of a
problem at the Watts Bar Nuclear plant in 1993.

Using plant-specific calculations, the Tennessee Valley Authority (TVA)
evaluated the service conditions for the ECCS throttle valves during post-LOCA
long-term recirculation (more than 100 days) at Sequoyah and Watts Bar.  The
licensee concluded that the valves might be operating under conditions that
could result in erosion-induced damage and eventual failure.  A Westinghouse
evaluation of the ECCS throttle valves confirmed TVA's analysis and 

9710280080.                                                            IN 97-76
                                                            October 30, 1997
                                                            Page 2 of 5


recommended that (1) the throttle valves be replaced with valves better able
to operate for extended periods under the analyzed conditions, (2) orifice
plates be installed in series with the throttle valves to reduce the pressure
drop across the valves, or (3) testing be performed to better define the wear
rate and determine appropriate corrective actions.  At both Sequoyah and
Watts Bar, orifice plates will be installed.

The Westinghouse evaluation identified two operational effects caused by
degradation of the ECCS throttle valves during a LOCA.  The first effect is
reduced throttling capability of the valves.  This occurs through erosion of
the valves as a consequence of cavitation engendered by high differential
pressure and flow across the valves.  As a result of the erosion, in
approximately 12 days the valves would be unable to adequately throttle ECCS
flow.  The second effect is increased pump flow in that valve erosion could
increase the valves' throttling area and result in increased pump flow.  In
this increased flow condition, the ECCS pumps could reach run out conditions,
resulting in pump damage.

Discussion

The subject throttle valves are located in ECCS flow paths for both the high-
head and intermediate-head safety injection pumps and are designed to balance
injection flows between loops while maintaining total injection flow within
normal parameters.

NRC requirements and guidance for the ECCS are provided in 10 CFR 50.46,
"Acceptance Criteria for Emergency Core Cooling Systems for Light-Water
Nuclear Power Reactors," and 10 CFR Part 50, Appendix A, "General Design
Criteria for Nuclear Power Plants," Criteria 33, 34, and 35.  Section 50.46
requires, in part, that "after any calculated successful initial operation of
the ECCS, the calculated core temperature shall be maintained at an acceptably
low value and decay heat shall be removed for the extended period of time
required by the long-lived radioactivity remaining in the core."  

To meet this requirement during a LOCA, the high-pressure ECCS injects water
into the reactor coolant system (RCS) cold legs to provide cooling flow to the
reactor core.  Following the initial injection phase, the ECCS pumps may be
aligned for cooling in the recirculation phase.  

During a LOCA, the ECCS may experience a large pressure drop across the
throttling valves, thereby causing cavitation and valve erosion.  As the
valves erode, their throttling capability is reduced.  This reduction, in
turn, increases ECCS pump flow.  If the ECCS pumps reach run out conditions as
a result of the continued valve erosion, the pumps may have to be stopped
before the plant can utilize the residual heat removal (RHR) system for core
cooling.

At Sequoyah and Watts Bar, the valves that could experience this type of
erosion are stainless steel, plug guided globe valves with stellite plugs and
seats.  Valves for this application are often supplied by the utility and/or
the architect engineer.  Thus, there may be other valve designs used in this
application.

Westinghouse analyzed long-term operability of throttle valves during both a
small-break LOCA (SBLOCA) and a large-break LOCA (LBLOCA).  During an SBLOCA,
the pressure   .                                                            IN 97-76
                                                            October 30, 1997
                                                            Page 3 of 5


drop across the valves, and the resulting valve erosion, is smaller than that
experienced during an LBLOCA. Relatively little valve erosion would occur
before the plant's emergency  operating procedures (EOPs) direct the operator
to secure the pumps.  As a result, if the pumps are secured in accordance with
the plant EOPs following an SBLOCA, minimal valve erosion will occur.  This
will allow the valves to perform their throttling function and will prevent
pump run out and subsequent pump damage during the SBLOCA.

However, for LBLOCAs, the rate and amount of valve erosion are greater because
of the greater differential pressure across the valves.  As a result, the
valves may erode to the point at which high- and intermediate-head pump flows
could reach run out conditions, thus necessitating the premature shutoff of
the pumps.

Another problem associated with erosion of this type of valve concerns
preventing boron precipitation in the RCS.  Without adequate flow through the
core,  boron will concentrate in the reactor vessel.  When the boron
concentration reaches the solubility limit, boron could plate out in the core
and inhibit flow.  If boron plate out occurs, the safety injection pumps and
the RHR pumps can be realigned to inject into the RCS hot legs to provide hot-
leg recirculation and prevent boron accumulation.  However, because the valves
in the recirculation lines for the safety injection pumps can also experience
the same valve erosion that occurs in the cold-leg injection line, the RHR
pumps may have to be used to provide hot-leg recirculation.  Depending on the
system design, in this mode of operation a single failure could occur which
could interrupt hot-leg recirculation flow, leading to possible boron
precipitation in the core.  This could occur if the single motor-operated
crossover valve fails.

Westinghouse, as part of TVA's Justification for Continued Operation (JCO),
analyzed four scenarios related to the effect of an LBLOCA on core cooling and
boron precipitation.  These scenarios constituted the worst-case challenges to
establishing and maintaining core cooling and hot-leg recirculation.  Each
scenario is discussed below.  

(1)   Large Cold-Leg Break With No Failure of the RHR Hot-Leg Motor-Operated
      Crossover Valve

      In this scenario, RHR flow would be realigned to the RCS hot legs
      approximately 12 hours after the LBLOCA (in accordance with the EOPs). 
      With the break in the cold leg and the RHR realigned to the hot leg, a
      flow path through the core is established and maintained while
      preventing boron buildup in the core.  In this scenario, flow from one
      RHR pump would be sufficient to maintain core cooling and prevent boron
      precipitation.

(2)   Large Cold-Leg Break With Failure of the RHR Hot-Leg Motor-Operated
      Crossover Valve

      In this scenario, an attempt would be made to realign RHR flow to the
      RCS hot legs approximately 12 hours after the LBLOCA (in accordance with
      the EOPs).  However, .                                                            IN 97-76
                                                            October 30, 1997
                                                            Page 4 of 5


      realignment would be unsuccessful due to the postulated failure of the
      motor-operated crossover valve.  RHR flow would then need to be
      realigned to the cold legs, resulting in cooling water bypassing the
      core and flowing out the break.  Boron concentration would increase,
      ultimately causing boron precipitation in the core.

(3)   Large Hot-Leg Break With No Failure of the RHR Hot-Leg Motor-Operated
      Crossover Valve

      In this scenario, RHR flow would be realigned to the RCS hot legs
      approximately 12 hours after the LBLOCA (in accordance with the EOPs). 
      Because the break is in the hot leg, injection flow would transit above
      the core and out the break without substantial mixing in the core.  For
      this scenario, adequate flow through the core is available if the
      hot-leg injection flow exceeds 3.3 times the boil off.  This requirement
      is met with one RHR pump.  

(4)   Large Hot-Leg Break With Failure of the RHR Hot-Leg Motor-Operated
      Crossover Valve

      In this scenario, an attempt would be made to realign RHR flow to the
      RCS hot legs approximately 12 hours after the LBLOCA (in accordance with
      EOPs).  However, realignment would be unsuccessful because of the
      postulated failure of the motor-operated crossover valve.  In this case,
      RHR flow would be realigned to the cold legs, thus allowing one RHR pump
      to provide adequate flow through the core and preventing boron buildup
      in the core.

Westinghouse evaluated these scenarios and determined that in each case, with
the exception of scenario 2, the flow from the RHR pumps was sufficient to
mitigate the accident.  In scenario 2, as there is only a single flow path for
hot-leg injection using an RHR pump, a failure of the motor-operated crossover
valve in this flow path could cause a loss of hot-leg injection during the
long-term recirculation cooling mode.  This scenario could result in blockage
of the reactor coolant flow path to the core because of boron precipitation
once the boron solubility limit was reached.  It should be noted that in
scenario 2, the review took credit for recirculation flow through the hot-leg
nozzle gaps in a forward flush path through the core in accordance with a
methodology described in Westinghouse NSAL-94-016, "Recriticality During LOCA
Hot-Leg Recirculation," dated July 25, 1994.  This methodology has not been
reviewed or approved by the NRC.  Therefore, depending on plant-specific
components and design, the conclusions may or may not be valid.  .                                                            IN 97-76
                                                            October 30, 1997
                                                            Page 5 of 5


This information notice requires no specific action or written response.  If
you have any questions about the information in this notice, please contact
one of the technical contacts listed below or the appropriate Office of
Nuclear Reactor Regulation (NRR) project manager.


                                    signed by

                                    Jack W. Roe, Acting Director
                                    Division of Reactor Program Management
                                    Office of Nuclear Reactor Regulation

Technical contacts:  Chu-Yu Liang, NRR
                     301-415-2878
                     E-mail:  [email protected]

                     William F. Burton, NRR
                     301-415-2853
                     E-mail:  [email protected]
```