ML20195H907
| ML20195H907 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 11/20/1998 |
| From: | Gwynn T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | Dugger C ENTERGY OPERATIONS, INC. |
| References | |
| 50-382-98-08, 50-382-98-8, EA-98-514, NUDOCS 9811240142 | |
| Download: ML20195H907 (8) | |
See also: IR 05000382/1998008
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WMGuy\\
UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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REGION IV
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611 RYAN PLAZA DRIVE, SUITE 400
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+9 *-****+0
ARLINGTON. TEXAS 760118064
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NOV201998
EA 98-514
Charles M. Dugger, Vice President
Operations - Waterford 3
Entergy Operations, Inc.
P.O. Box B
Killona, Louisiana 70066
SUBJECT: NRC INSPECTION REPORT 50-382/98-08 AND NOTICE OF VIOLATION
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Dear Mr. Dugger:
Thank you for your letter of July 1,1998, in response to our June 1,1998, letter and Notice of
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Violation. We have reviewed your reply to Violations 50-382/9808-01 and -04 and find it
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responsive to the concerns raised in these violations. Your reply to Violation 50-382/9808-04
stated that you disagreed with the NRC's position that this violation was similar to a violation
cited in NRC Inspection R,eport 50-382/98-06. We agree with your position that the two
violations are not similar as a result of further review of the details of the violations. It appears
that the timeliness of the actions you implemented to address the issues identified in Violation
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50-382/9808-04 was appropriate. We will review the implementation of your corrective actions
for these two violations during a future inspection to determine that full compliance has been
achieved and will be maintained.
Your response to Violation 50-382/9808-03 denied that a violation of Criterion 111 of Appendix B
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to 10 CFR Part 50 occurred. The position provided in your response basically stated that no
operational limit specified for safety-related structures, systems, and components in procedures
should be considered as part of the design basis. Your conclusion appears to be based on
your assertion that the acceptance criteria provided in procedures for testing of structures,
systems, and components may be exceeded, pending the completion of an engineering
evaluation of the out-of-specification data to determine if the bounding limits established by the
accident analysis were not exceeded.
We disagree with your stated position on Violation 50-382/9808-03 for the following reasons:
Your response discusses the definition of design basis, as defined in 10 CFR 650.2,
and discusses why _the violation is not valid because the basis for the violation does not
fall within the definition provided in 650.2. Criterion lll of Append;x B to 10 CFR Part 50
states, in part, ' Measures shall be established to assure that applicable regulatory
requirements and the design basis, as defined in $50.2 and as specified in the license
application, for those structures, systems, and components to which this appendix
applies are correctly translated into specifications, drawings, procedures, and
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instructions." From this statement, it is apparent that the design basis has been
established by Criterion lll to include: (1) the design basis as defined in G50.2 and (2)
the design basis as specified in the license application. During review of your response,
9811240142 981120
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ADOCK 05000382
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Entergy Operations, Inc.
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it was noted that you did not address both aspects of design basis in that the response
did not discuss the license application aspects.
Criterion 111 also states, in part, that, " Design control measures shall be applied to items
such as the following: . . . delineation of acceptance criteria for inspections and tests." It
is apparent from this statement that the intent of Criterion ill was that acceptance criteria
be provided in procedures based on the design basis of the system. Your response did
not address this aspect of the requirements of Criterion 111.
As discussed in 50.34, the license application includes the Technical Specifica,tions,
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Final Safety Analysis RepM, and many other types of documents. Since Criterion 111
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includes these documents as being a part of the applicable design basis, then the data
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contained in these documents represent design basis information and are required to be
correctly translated into specifications, drawings, procedures, and instructions. In
addition, Criterion 111 specifically requires that criteria be provided in testing procedures.
Violation 50-382/9808-03 was cited because you failed to translate design basis
requirements (engineered safety features system response times) into the appropriate
procedures, as required by Criterion Ill.
In your response, you stated that the significance of the specific example (emergency
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feedwater system tesponse time) was minor since the value of the error was of no
consequence. Your response also stated that the two other examples (involving two
other safety-related systems) provided in the subject inspection were also minor. . Your
stated basis for this conclusion is that testing had previously confirmed that the
measured parameters were within the established acceptance criteria. We agree with
this observation. However, the safety significance of this violation is not based on the
quantitative results of system testing, rather it is based on the number of identified
errors in the established testing acceptance criteria for three different safety-related
systems. This is an indication of a potential generic problem in maintaining design basis
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information in procedures.
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Based on the above discussion. we have determined that Violation 50-382/9808-03 remains
valid as previously issued in the subject inspection report.
It is requested that you reply, within 30 days of the date of this letter, with a written response
discussing the corrective actions you plan to implement to address the root cause(s) of
Violation 50-382/9808-03.
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Should you have any questions regarding this letter, we will be pleased to discuss them with
you.
Sincer
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homas P. Gwynn
irec or
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Division of Reacto Proj ts
Docket No.:
50-382
. License No.: NPF-38
cc:
Executive Vice President and
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Chief Operating Officer
Entergy Operations, Inc.
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P.O. Box 31995
Jackson, Mississippi 39286-1995
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Vice President, Operations Support
Entergy Operations, Inc.
P.O. Box 31995
Jackson, Mississippi 39286-1995
Wise, Carter, Child & Caraway
P.O. Box 651
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Jackson, Mississippi 39205
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General Manager, Plant Operations
Waterford 3 SES
Entergy Operations, Inc.
. P.O. Box B
Killona, Louisiana 70066
Manager- Licensing Manager
Waterford 3 SES
Entergy Operations, Inc.
P.O. Box B
Killona, Louisiana 70066
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Chairman
Louisiana Public Service Commission
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One American Place, Suite 1630
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Baton Rouge, Louisiana 70825-1697
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Entergy Operations, Inc.
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Director, Nuclear Safety &
Regulatory Affairs
Waterford 3 SES
Entergy Operations, Inc.
- P.O. Box B ~
Killona, Louisiana 70066
William H. Spell, Administrator
Louisiana Radiation Protection Division
P.O. Box 82135
Baton Rouge, Louisiana 70884-2135
Parish President
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St. Charles Parish -
P.O. Box 302
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Hahnville, Louisiana 70057
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Winston & Strawn
1400 L Street, N.W.
Washington, D.C. 20005-3502
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Project Engineer (DRP/D)
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perations, Inc.
Killona.LA 70066
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Tel 504 739 6242
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Early C. Ewing,ill
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le Safety & Regulatory Affairs
Waterford 3
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A4.05
PR
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July 1,1998 -
U.S. Nuclear Regulatory Commission
ATTN: Document Control Desk
Washington, D.C. 20555
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Subject:
Waterford 3 SES
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Docket No. 50-382
License No. NPF-38
NRC Inspection Report 98-08
_ Reply to Notice of Violation
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1: In accordance with 10CFR2.201, Entergy Operations, Inc. hereby submits in
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Attachment 1 the response to Violations 9808-01,9808-03 and 9808-04 identified in
Enclosure 1 of the subject inspection Report. On June 29,1998, an extension of the
original 30-day response date until July 10,1998, was granted to Waterford 3 by Mr.
P. Harrell, NRC Region IV.
Waterford 3 agrees with the NRC's assessment of Violations 9808-01 and 9808-04.
However, we do not agree with the characterization of Violation 9808-04 as an
example of narrowly focused scoping of a" problem, similar to the issue with
diaphragm valves discussed in NRC Inspection Report 50-382/98-06. This is-
discussed in Attachment 1.
With regards to Violation 9808-03, we have performed a critical examination of the
inspection report and do not find this to constitute a condition outside of the plant's
design basis. Additional details are provided in Attachment 1.
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NRC Inspection Report 98-08;
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Replyto Notice of Violation '
Page 2
July 1,1998
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If you have any questions ~conceming this response, please contact
me at (504) 739-6242.
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Very truly yours, .
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E.C. Ewing
Director
Nuclear Safety & Regulatory Affairs
. I ECE/BVR/rtk
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cc:
E.W. Merschoff (NRC Region IV), C.P. Patel (NRC-NRR),
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J. Smith, N.S. Reynolds, NRC Resident inspectors Office
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ATTACHMENT 1
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ENTERGY OPERATIONS. INC. RESPONSE TO THE VIOLATION IDENTIFIED IN
ENCLOSURE 1 OF INSPECTION REPORT 98-08
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VIOLATION NO. 50-382/9808-01
Technical Specification 6.8.1.a requires, in part, that written procedures be
established covering the applicable procedures recommended in Appendix A of
Regulatory Guide 1.33, Revision 2, Febfuary 1978. Section 3 of Appendix A requires
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that the licensee have procedures for the operation of emergency power sources.
Contrary to the above, the licensee failed to establish adequate procedures for the
operation of the emergency diesel generators when in a no or low electrical loading
condition in that no specific guidance was provided for operational restrictions on the
operation of the emergency diesel generators in a low or no load condition.
This is a Severity Level IV violation (Supplement 1) (50-382/9808-01).
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RESPONSE
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Summary of Entemy Operations, Inc. Position
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Entergy Operations Inc. has carefully evaluated the information in Violation 9808-01
and agrees with the NRC's assessment that operations procedure OP-009-002
regarding diesel generator operation at low or no load conditions did not provide
appropriate guidance to the operations staff.
Reason for the Violation
The reason for the inadequate procedure, regarding diesel generator operation at
low or no load conditions, is an inadequate review of the content of Operating
Procedure, OP-009-002, Emergency Diesel Generator (EDG) due, in part, to vague
guidance provided by the emergency diesel generator vendor. In severallocations
throughout the procedure, it was stated that the EDG should not be operated for an
extended period of time unloaded. The procedure also stated that the fuelinjection
pump temperatures should be checked periodically if the EDG is operating at an
unloaded condition for an extended period of time. If any fuelinjection pump gets
too hot to comfortably hold your hand on, then load or secure the EDG. The length
of time to operate the EDG unloaded as defined by the phrase " extended period of
time" was unclear in the procedural instructions provided to operations personnel.
However, because the EDG vendor manual did not provide a defined period of time
to run the EDG in an unloaded configuration, the phrase " extended period of time"
was used in the original development of OP-009-002 to provide guidance to the
operator to limit operating the diesel at unloaded conditions. It should be noted that
several recent inquiries have been made to the vendor regarding time limits for
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running the diesel unloaded, and, the vendor has stated that this information is
unavailable.
A contributing cause for this occurrence is a lack of questioning attitude by
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operations personnelin that the vagueness of the phras'e " extended period of time"
and the uncertainty of the actual temperature of the injection pump to begin loading
or securing the EDG was not questioned. The procedure required loading or
securing the EDG when the injection pump gets too hot to comfortably hold your
hand on the pump. This " temperature value" is subjective and was not quantified.
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Corrective Steps That Have Been Taken and the Results Achieved
Operation Procedure OP-009-002, Emergency Diesel Generator, was revised on
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May 17,1998, to clarify expectations of running the EDG unloaded or at low load
conditions. The issue was investigated by Waterford 3 Systems Engineering and
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discussions were heid with the EDG vendor. it was determined that a concern for
overheating the fuel injection pumps at no or low loads does not exist. Thus, the
phrase " extended period of time" was removed throughout the procedure and a
. requirement was added to minimize operation at no or low load conditions. The
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iintent,of the new requirement is to prompt the operators to secure the EDG if they
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jdetermine the dieselis not needed.
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iln additiori, the requirement to load or secure the diesel when the injection pump
gets too hot to comfortably hold your hand on the pump was determined not
necessary by the EDG vendor and Waterford 3 System Engineering. Thus, it was
removed from the procedure.
On May 13,1998, the Operations Superintendent discussed with operations
personnel the need to address vagueness in procedures and to take the appropriate
steps to correct the deficiency. A memo was issued to operations personnel on May
15,1998, to reiterate these expectations.
Corrective Steps Which Will Be Taken to Avoid Further Violations
The above corrective actions are adequate to avoid further violations.
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Date When Full Compliance Will Be Achieved
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Waterford 3 is in full compliance.
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VIOLATION NO. 50-382/9808-03
1b CFR Part 50, Appendix B, Criterioi lli states, in part, that measures shall be
established to assure that applicable regulatory requirements and the design basis
are correctly translated into procedures and instructions.
Contrary to the above, the licensee failed to correctly translate applicable regulatory
requirements and the design basis into procedures and instructions in that the
response time acceptance criteria contained in surveillance test procedures for the
emergency feedwater system, containment fan coolers, and high pressure and low
pressure safety injection systems did nof ensure that the requirements of the
licensing basis were met.
This is a Severity Level IV violation (Supplement 1) (50-382/9808-03).
RESPONSE
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Summary of Enteray Operations, Inc.. Position
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- Entergy Operations Inc. has carefully evaluated the information in Violation 9808-03
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- Control," did not occur.
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Basis for Enterav Operations, Inc., Position
Nuclear power plants are designed to operate according to a design basis.10 CFR 50.2 defines design basis as information which identifies the specific functions to be
performed by a structure, system, or component (SSC) of a facility, and the specific
values or ranges of values chosen for controlling parameters as reference bounds for
- design.10 CFR 50.2 also states, in part, that design basis values are requirements
derived from analysis of the effects of a postulated accident for which a SSC must
meet its functional goals. Based on this definition for design basis, Waterford
considers the following position on conditions outside the design basis of the plant to
be appropriate.
The functional goal of a safety-related SSC in a nuclear power plant is to perform its
required safety function as assumed in the applicable accident analyses. The results
of the accident analyses serve as the reference bound for overall plant design to
ensure that the plant is capable of mitigating the consequences of a design basis
accident. In order to assure the health and safety of the public, some analysis
results (i.e., cladding temperature, cladding oxidation, etc.) coincide with regulatory
requirements and are affected by a range of input values. EFW response time is an
example of such an input value. While a failure to meet the acceptance criteria
during testing of a SSC would constitute a degraded or nonconforming condition,
such a condition is not necessarily outside the desian basis of the plant. If changes
to a model input significantly affect the ability of the SSC to perform its intended
safety function, as determined through analysis, a condition would exist which is
outside of the plant's design basis. For this purpose, a significant effect is one wF 'h
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results in exceeding regulatory requirements. This position is consistent with similar
regulations, such as 10 CFR 50.46(a)(3)(i), which addresses significant changes in
the application of Emergency Core Cooling System models.
As a result of the above position, transients are typically simulated over the time
period that the results have a potential to approach the design or regulatory criteria.
With respect to the EFW turbine driven pump, results of UFSAR Chapter 15.2
accident analyses occur before EFW flow is initia.ted. Thus, an increase in response
time of the magnitude seen here does not have the potential to affect the results of
any of the associated analyses. Based on this information, the condition would not
result in a condition that is outside of th5 plant's design basis.
Wateiford doe.s believe, however, that periodic testing of the EFW turbine driven
pump should be performed in accordance with adequate written procedures, as
required by TS 6.8.1.c. Therefore, Waterford has developed corrective actions to
address deficiencies related to operations procedure OP-903-047, " Emergency
Feedwater Actuation Signal Test."
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Violation 9808-03 also refers to the Containment Fan Coolers, High Pressure and
Low Pressure Safety injection Systems. Based on an investigation of the issue,
> Waterford concluded that the actual surveillance data for these systems met the
acceptance criteria and preserved the values given in the Technical Requirements
Manual. Thus, no viciation occurred with respect to these systems. Furthermore,
NRC inspection Report 98-08 states:
As resuit of this discovery,.the licensee reviewed TRM Table 3.3-5
to identify any other inconsistencies of this type. Two other similar
conditions were identified and documented as follows:
CR 98-0545
Start Response Time for Containment Fan
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Coolers
CR 98-0558
Start Response Times for the High
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Pressure Safety injection (HPSI) and Low
Pressure Safety injection (LPSI) Systems
in both these cases, the acceptance criteria contained in the
surveillance procedures could possibly result in the response times
assumed in the FSAR and the TRM requirements being exceeded.
However, a review of the actual test results indicated that the
response time was within the requirements for both these cases.
Because of this, the containment fan coolers and HPSI and LPSI
pumps were not declared inoperable.
The inspectors reviewed the above referenced documents and
discussed these issues with licensee personnel. Based on these
reviews, it appeared that the actions taken by the licensee were
appropriate upon discovery of the inconsistencies.
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. Corrective Steps That Have Been Taken and the Results Achieved
Design Engineering performed an evaluation to determine if other pumps that
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receive an ESFAS actuation have the same potential condition. These pumps
were identified and discussed in Condition Reports (CRs) 98-0545 and 98-0558.
Operations revised procedure OP-903-047," Emergency Feedwater Actuation
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Signal Test," to ensure surveillances contain an allowance for differences in
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automatic actuation from a setpointqn the Plant Protection System (PPS) and
manual actuation from the " Initiate" test button.
Corrective Steps Which Will Be Taken to Avoid Further Violations
A root cause analysis is being performed to address the causes of this condition.
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Date When Full Compliance Will Be Achieved
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VIOLATION NO. 50-382/9808-04
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Technical Specification 4.8.4.2.b.1 requires, in part, that all thermal overload devices
for motor-operated valves, which are not bypassed, be calibrated at least once every
6 years.
Contrary to the above, the licensee failed to calibrate all thermal overload devices
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every 6 years that the overload devices for motor-operated valves MS-401 A(B), SI-
135A(B), SI-125A(B), and SI-412A(B) had not calibrated in the previous 6 years.
This is a Severity Level IV violation (Suphlement 1) (50-382/9808-04).
RESPONSE
Summary of Enteray Operations. Inc. Position
EOl has carefully evaluated the information in Violation 9808-04 and agrees with the
NRC's assessment of this condition as a violation of TS 4.8.4.2.b. However, we do
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e not agree that this is an example of narrowly focused scoping of a problem, sim lar to
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l j the issue,with the diaphragm valves discussed in NRC Inspection Report 50-382/98-
06.
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. With respect to the diaphragm valves issue, Waterford incorrectly limited the
population of potentially affected valves to those modified per a Design Change (DC-
3211). No effort was made by Waterford to inspect the stop nuts on other similar
valves to determine the full scope of a stop nut misadjustment problem.
Contrary to the diaphragm valve issue, the thermal overload concerns were identified
by Waterford and an ongoing investigation of other motor operated valves with
thermal overload protection and/or bypass devices was being conducted by
Waterford at the time the issue was discussed with the NRC. Although all of the
potentially affected valves were not immediately identified, the fact that they were
eventually identified through an open item in the Waterford Corrective Action
Program indicates a broad and comprehensive approach.
Reason for the Violation
The root cause of this violation was determined to be an inadequate evaluation of
what valves should have been incorporated in Technical Requirements Manual
(TRM) Table 3.8-2. TRM Table 3.8-2 was originally included in Waterford Technical Specification (TS) 3.8.4.2 and was later moved into the TRM. The table lists motor
operated valves which have thermal overload protection and/or bypass devices.
In 1980. Crosby Valve Division (Report No. 4093) conducted tests which resulted in
the replacement of the air operators in MS-401 A(B) with motor operators. As a
result of this change, MS-401 A(B) should have been added to Table 3.8-2.
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However, this need was not recognized and the valves were never added to the
table.
Unlike MS-401 A(B), Safety injection Valves SI-125 A(B), Sl-135 A(B) and SI-412
A(B) were always motor operated valves. These valves are associated with the
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Shutdown Cooling System and appear to have been omitted from Table 3.8-2 due to
an inadequate evaluation at the time the table was originally developed. This was
apparent from the inclusion of similar valves, SI-4,15 A(B). Although SI-125 A(B), SI-
135 A(B) and SI-412 do not receive an ESFAS signal they are safety related and
have always possessed the characteristics for inclusion in Table 3.8-2.
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Additional information and details concerning this condition were included in
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Licensee Event Report (LER) 98-005-00, which was submitted to the NRC on April
13,1998.
Corrective Steps That Have Been Taken and the Results Achieved
The thermal overloads for all 8 motor-operated valves were removed and
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successfully tested in accordance with TS surveillance requirement 4.8.4.2.b.
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Desijn Engineering reviewed the safety related motor-operated valves with
then651 overloads used in safety systems.
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Design Engineering added valves MS-401 A(B), SI-135A(B), SI-125A(B) and SI-
412A(B) to TRM table 3.8-2.
Maintenance Procedure ME-003-410, " Motor-Operated Valve Thermal Overload
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Channel Calibration," was revised (Rev. 8) to include SI-125 A(B), SI-135 A(B),
and SI-412 A(B). Furthermore, the following repetitive tasks were initiated to
ensure testing of these valves in accordance with ME-003-410:
Repetitive Task
Valve
Number
022564
SI-135A
022565
Sl-135B
022562
SI-125A
022563
SI-125B
016410
SI-412A
022566
SI-4128
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' Corrective Steos'Which Will Be Taken to Avoid Further Viola'tions
A review of other TS revisions made in accordance with Generic Letter
(GL) 91-08 will be performed to ensure the changes specified in GL 91-08 were
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adequately evaluated.
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Date When Full Compliance Will Be Achieved
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The action to review other TS revisions'inade in accordance with Generic Letter (GL)
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91-08 will be completed August 27,1998.
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