United States Nuclear Regulatory Commission - Protecting People and the Environment

ACCESSION #: 9501270196

                                                TELEPHONE (815) 727-2600
                                                    POST OFFICE BOX 3339
                                                       JOLIET, IL  60434

CRANE                      NUCLEAR OPERATIONS


August 30, 1994

Nuclear Regulatory Commission
Document Control Desk
Washington, DC  20555


REFERENCE:      a.)  Crane Letter dated August 23, 1994
                b.)  Northern States Power Letter to the
                     NRC dated April 18, 1994


This letter will serve to advise you of our continuing investigation
concerning Crane design Gate Valves involved in the January, 1994
incident at the Prairie Island Nuclear Generating Plant.  Details of
that incident were contained in the reference (b) correspondence, a
copy of which as attached.

We have been in constant communication with Westinghouse Electric's
NATD Engineering Technology Department concerning their December, 1969,
P.O. 546-CAK-116878 BN on which the valves in question were purchased
from Crane Company.  In turn, Westinghouse supplied the valves to a
number of Nuclear Utilities, records for which have been retrieved from
Crane archives.

On August 23, 1994, a letter was issued by Crane to those Utilities to
whom valves were supplied.  A sample copy of the letter is attached for
your reference and a total list of Utilities to whom the letter was
sent is shown on the attached page.

At this point, we consider the incident a potentially reportable
condition based on the NSP evaluation at their plant.  Upon receipt of
additional input from the balance of the plant, we will assess the
situation to determine a subsequent course of action.  Historical
records do not contain information addressing the total number of
valves of this configuration manufactured by Crane, nor their
application throughout the Industry.

Page 2

As additional information becomes available, we will advise.  You may
direct any questions to the writer or to our Manager of Developmental
Engineering, Mr. Bruce Harry.  My direct line is 815-740-7597 and Bruce
can be reached at 815-740-7570.


Ronald F. Hornyak
Manager QA/Support Engineering


cc:   J. Carlson
      F. Bisesto
      B. Harry
      K. Hutchinson
      H. Sandner (Westinghouse)


                                                TELEPHONE (815) 727-2600
                                                    POST OFFICE BOX 3339
                                                       JOLIET, IL  60434

CRANE                      NUCLEAR OPERATIONS


August 30, 1994

Beaver Valley Power Station
Duquesne Light Company
P.O. Box 4
Shippingport, PA  15077


In January of this year a containment isolation valve at the Prairie
Island Nuclear Station was damaged as a result of operational problems
involving maintenance procedures.  Details of the event have been
reported to the NRC via Prairie Island's Docket Nos. 50-282 and 50-306
and associated NRC Inspection Report Nos. 282/94002 and 306/94002.

The valve in question was a Crane 10" Figure 63174 and during the
course of the investigation, it was discovered that, contrary to
information shown on the valve assembly drawing, shims were installed
between the yoke and the adapter plate.  The extraordinary conditions
experienced by the valve during the event resulted in failure of the
welds joining the yoke and adapter plate and more catastrophic failure
to the other valve parts.

Crane assisted the Utility in evaluating additional valves of the same
design for evidence of the existence of shims and for the adequacy of
the attachment welds where shims did exist.  It was concluded that a
substantial safety hazard did not exist at the Prairie Island Plant,
based on the walk-down that was performed and additional analytical
evaluations of the as-built valve characteristics.

Since that incident, Crane has conducted additional evaluations,
including a search of archives stored in a Crane facility in
Pennsylvania.  The valves at the Prairie Island plant were part of an
order produced by Crane for Westinghouse Electric in the early 1970's. 
Although the records retrieved contained no information to support the
addition of shims to the design, the records indicate that valves of
the Figure 63174 design were supplied to you on that same Westinghouse

Crane requests your cooperation in determining if those valves are in
your system.  We will be glad to assist you in evaluating the condition
of the hardware and in performing any analytical evaluations to
determine if safety hazards exist.  Valves of the 63174 design were
supplied in 10" and 12" sizes.

You have been contacted based upon your affiliation with the Motor
Operated Valve User's Group and its Member Roster.  We appreciate your
cooperation in channeling this communication to the appropriate
personnel at your facility.

Your questions can be directed to the writer or to our Manager of
Development Engineering, Mr. Bruce Harry (who performed the initial on-
site evaluation at Prairie Island) or Mr. David Dwyer, our Analytical
Project Engineer, who has paid subsequent visits to the Prairie Island
plant.  Our general number in Joliet is 815-727-2600.

Very Truly Yours,

R.F. Hornyak

Manager QA/Support Engineering


cc:   J. Carlson
      K. Hutchinson
      B. Harry
      D. Dwyer


         Names / Address of Recipients of August 23, 1994 Letter

Beaver Valley Power Station         Kewaunee Nuclear Power Plant
SEB-3                               Wisconsin Public Service Corp.
Duquesne Light Company              N 490 Hwy 42
PO Box 4                            Kewaunee, WI  54216
Shippingport, PA  15077

ATTN:      Mr. Neil Morrison        ATTN:     Mr. Larry I. Limberg
           Senior Engineer                    Maintenance Engineer

Diablo Canyon Power Plant           MC N-50
Pacific Gas & Electric Co.          Public Service Electric & Gas
PO Box  56                          PO Box 236
Avila Beach, CA  93424              Hancocks Bridge, NJ  08038

ATTN:      Mr. Don Bauer            ATTN:     Mr. Robert S. Lewis
                                              Senior Staff Engineer

Corporate Office
Tennessee Valley Authority
1101 Market St.
Chattanooga, TN  37402-2801

ATTN:      Mr. Richard G. Simmons
           Program Manager, Valves

Innsbrook Technical Center
Virginia Power Company
5000 Dominion Blvd.
Glen Allen, VA  23060

ATTN:      Ms. Pamela E. Detine
           Project Engineer

NSP                            Northern States Power Company

                               Prairie Island Nuclear Generating Plant

                               1717 Wakonade Dr. East
                               Welch, Minnesota 55089

April 18, 1994                           10 CFR Part 21
                                         Appendix C

US Nuclear Regulatory Commission
Attn:  Document Control Desk
Washington, DC  20555

                 Docket Nos. 50-282   License Nos. DPR-42
                             50-306                DPR-60

Reply to a Notice of Violation
NRC Inspection Report Nos. 282/94002(DRP) and 306/94002(DRP)
Procedural Deficiency Allowing Damage to a Safety-related Valve

Your letter of March 18, 1994 transmitted the subject inspection report
and violation notice which required a 30 day response.  Attached is our

In our response, we have made new NRC commitments which are identified
as such in the attachment as the statements which are in italics.

If you have any questions regarding this response, please contact Jack
Leveille (612-388-1121. extension 4562).

E L Watzl
General Manager
Prairie Island Site

cc:   Regional Administrator III, NRC
      Senior President Inspector, NRC
      NRR Project Manager, NRC
      J. E. Silberg


                          RESPONSE TO VIOLATION


      Criterion V of 10 CFR 50, Appendix B, requires that activities
      affecting quality be prescribed by documented instructions,
      procedures, or drawings, of a type appropriate to the
      circumstances and shall be accomplished in accordance with these
      instructions, procedures, or drawings.

      Contrary to the above, on January 24, 1994, extensive damage to
      the normally closed outboard containment isolation valve
      (MV-32181) between the containment building sump and the suction
      of No. 22 residual heat removal pump occurred when the
      electrician, stroking the valve locally from the motor control
      center, continually depressed the close contactor instead of
      depressing the open contactor.  The maintenance procedure used to
      locally stroke the valve was not appropriate to the circumstances
      in that: 1) the procedure did not require that direct
      communications be established between the electrician and
      operators during stroking of the valve; 2) expected values for
      motor current draw were not included; 3) the method of making up
      the contactor was not specified (i.e. the open contactor did not
      have to be continually depressed in order to operate the valve):
      and 4) the procedure required errorless human performance because
      of the absence of actuator protective features.

      This is a Severity Level IV Violation (Supplement I).


We consider this failure to be significant since a risk-significant
component failed, albeit not while performing in the accident
functional mode.

Although the work procedure for cycling the valve had the deficiencies
noted in the violation, it should be noted that the plant electricians
have cycled motor valves locally at the breaker using the contactors
numerous times as part of motor-operated valve (MOV) testing efforts. 
Such cycling activities have occurred since the 1980's.  In an effort
to reduce the likelihood of error, the motor valve and system engineers
initiated the procedure referenced in the violation.  In preparing the
procedure the engineer had spoken with electrical maintenance personnel
experienced in motor operated valve testing to ensure that the
procedure would be appropriate.  Based on this input, the procedure was
developed for cycling MV-32181.

MV-32181 was the third valve to be cycled in response to a NRC
identified concern regarding the potential for pressure locking.  The
person assigned to perform the task was a journeyman electrician. 
Previously in January, two different electricians had used the same
procedure satisfactorily on the two Unit 1 valves.


                                                             Page 2 of 5

Reason for the Violation

Two root cause analyses were initiated - one to determine the root
cause for the event and the second to determine the valve failure root
cause (this is discussed in the "Corrective Steps Taken and Results
Achieved" section).  The event root cause analysis was performed by the
on-site Error Reduction Task Force (ERTF) and is documented in ERTF
Report 94-01.

The ERTF report identified two primary causes, two secondary causes,
and two additional possible causes for the event.  The primary cause
was determined to be human error.  The causes were:

      Primary causes - Human Error

           1.   Self-checking was not applied to verify that the choice
                of contactor was correct, or that the intended action
                was correct.  The electrician depressed the wrong

           2.   The electrician did not have the proper information at
                the job site to verify whether the valve's circuit was
                seal-in or not.  Without this information, the
                electrician pressed and held the contactor to ensure
                the valve would go open for the required 30 seconds. 
                Holding the contactor in bypasses the torque switch

      Secondary causes - Ergonomics

           1.   The open and close contactors in the MCC breaker
                cubicle were not labeled.

           2.   The work request was somewhat generic in that it did
                not specify the expected current draw.  Also, the work
                request did not contain instructions to push and
                release the contactor, nor did it mention the seal-in

      Possible contributing causes

           1.   No communications were established between the MCC
                breaker cubicle and the motor valve.

           2.   Consequences of potential error were not discussed before
                starting the work.

Corrective Steps Taken and Results Achieved

The affected valve was repaired and restored to service on January 26,
1994.  Prior to cycling additional motor operated valves (e.g., MV-32180,


                                                             Page 3 of 5

Sump B to 21 RHR Pump) by this method, the causes of the event were
identified and corrective actions to prevent a similar event were
discussed with the plant electricians.  A more detailed procedure was used
for the cycling of the next valve, MV-32180, which was cycled successfully
on February 4, 1994.

Prairie Island has developed a videotape intended to emphasize
self-checking.  This video, "Right from the Start", has now been viewed
by some, but not all, of the plant staff, including the electricians.

A method was implemented to label the open and close contactors in motor
valve MCC breaker cubicles.  To date, the contactors in 142 MCC breaker
cubicles have been labeled.

On March 24, 1994, temporary memos (94-24 and 94-25) were issued to both
units' quarterly surveillance procedures, SP1089 and SP2089 (Residual Heat
Removal Pumps and Suction Valves from the Refueling Water Storage Tank),
respectively.  These temporary memos are refinements of the procedure used
for the MV-32180 cycling.  These refinements were developed during a
post-event evaluation by those involved in the event.  These procedure
changes involved cycling the Sump B valves locally for potential pressure
locking.  The procedure specified the following additional
information/requirements beyond those specified in the original work

      (1)  precaution to self-check,

      (2)  perform a pre-job briefing,

      (3)  use of headset communication between the electrician at the MCC
           breaker cubicle and the operator at the valve,

      (4)  identification of the expected full load and nameplate locked
           rotor amperage,

      (5)  caution that the contactor need not be held in, but only
           momentarily depressed since it is a seal-in circuit, and

      (6)  verification that the open and closed contactors are labeled.

The MOV testing engineers were advised to ensure adequate instructions are
provided for those cases where local operation of an MOV is required. 
They will evaluate their procedures for necessary changes.

A Safety Evaluation revision provides short term justification of the
operability of the Sump B valves based on the calculational methodology
of required opening thrusts under pressure locking condition.  These
valves will no longer be cycled for pressure locking concerns.  A
modification is being prepared to modify the valves to prevent pressure

The Equipment Failure Root Cause evaluation was initiated by the


                                                             Page 4 of 5

staff with the assistance of the valve manufacturer (Crane/Aloyco), an
independent engineering firm (Altran) experienced in failure analysis of
this type, and the corporate Materials and Special Processes department. 
The vendor analyses are still in progress.

Mechanically the weld failed at the point where the yoke arm was attached
to the actuator adapter plate.  This failure location was not predicted
by the valve manufacturer's weak link analysis.  Rather, failure was
anticipated in the yoke.  The failure root cause analysis has determined
that the following additional factors contributed to the premature

      (1)  A shim was installed between the yoke and adapter plate that
           resulted in a lower failure stress than design,

      (2)  The shim was not indicated on the fabrication drawings,

      (3)  The shim was not included in the design calculations, and

      (4)  The weld that broke was poor quality as indicated by
           less-than-design fusion.

Preliminary data from Crane and Altran indicate that the as-built values
for the weldments were less than the original Westinghouse specification
design values.  These additional factors are apparently due to inadequate
Crane quality assurance and controls and inadequate Westinghouse
oversight.  Crane was the manufacturer and Westinghouse was the supplier.

We have determined that the as-built valve characteristics do not
constitute a substantial safety hazard for the Prairie Island plant
application.  However, the existence of the shims in valves in different
applications in other plants may present a substantial safety hazard.  We
have notified Westinghouse that they may need to perform an evaluation for
10 CFR Part 21 reporting purposes.

Although the torque switch was bypassed due to the actions of the
electrician, the torque switch settings could have been set at a lower
setting since the plant design differential pressure is 46 psid rather
than the generic 700 psid design differential pressure.

The discovery of the shim led to expansion of the investigation to all
other Crane valves of this type used in safety-related applications. 
These valves include:

Containment Sump B to RHR:     12"  32075, 32076, 32077, 32078  (Unit 1)
                                    32178, 32179, 32180, 32181  (Unit 2)

RWST to RHR Pumps:             10"  32084, 32085                (Unit 1)
                                    32187, 32188                (Unit 2)

The four RWST to RHR Pump valves and the other three outside Sump B to RHR


                                                             Page 5 of 5

valves were inspected for presence of a shim and weld quality.  The
results of the inspections were satisfactory.  The four inside Sump B to
RHR valves were not inspected at this time since a unit shutdown would be
needed and it is believed that inspection of the remaining valves provides
a reasonable expectation that the valves inside the containment boundary
are not different.

The torque switch setting for MV-32181 has been adjusted downward to
correlate closer to the 46 psid plant design than the higher generic
design differential pressure.  This change had been planned prior to the
event occurrence.

A further investigation of the maintenance history showed that one of the
Unit 1 RWST to RHR pump valves failed in 1975.  At that time those four
valves were evaluated and rewelded as appropriate.

Corrective Steps That Will Be Taken To Avoid Further Violations

The need for a revision of the plant maintenance procedure writers' guide
will be reevaluated, by June 1, 1994, in light of the observations made
during the evaluation of this event.

A Maintenance section procedure will be completed, by June 1, 1994, that
describes in detail the method for cycling a motor valve locally by using
open and close contactors since this approach is used in the MOV testing

The remaining containment sump B valves will be inspected for presence of
the shim and weld quality and the torque switch settings will be
readjusted to correlate with the lower plant design differential pressure,
by July 1, 1994 for Unit 1 and July 1, 1995 for Unit 2.

The videotape "Right from the Start" has been incorporated into the
General Employee Training re-qualification program, which is presented to
all personnel badged for access to the plant.

Date When Full Compliance Will Be Achieved

Full compliance has been achieved.


Page Last Reviewed/Updated Monday, January 27, 2020