ML20207P264
| ML20207P264 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 01/05/1987 |
| From: | Brey H PUBLIC SERVICE CO. OF COLORADO |
| To: | Gagliardo J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| References | |
| P-87001, NUDOCS 8701150323 | |
| Download: ML20207P264 (9) | |
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2420 W. 26th Avenue, Suite 1000, Denver, Colorado 80211 January 5, 1986 Fort St. Vrain Unit No. 1 P-87001 Regional Administrator Region IV U. S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000
$h5((h Arlington, Texas 76011 1
i'I JAN 81987 Attention: Mr. J. E. Gagliardo, Chief
. t Reactor Projects Branch Ll h,
Docket No. 50-267
SUBJECT:
I&E Inspection 86-25
REFERENCE:
(1) NRC Letter, Gagliardo to Williams dated November 4,1986(G-86584)
(2) PSC Letter, Brey to Gagliardo dated December 4, 1986 (P-86656)
Dear Mr. Gagliardo:
This letter is in response to the Notices of Violation received as a result of inspections conducted at Fort St. Vrain during the period of August 16 through September 30, 1986.
An extension to the response date contained in Reference 1 was granted as indicated by Reference 2.
1.
Violation of Shift Turnover Procedure Criterion V of Appendix B to 10 CFR Part 50, and the licensee's approved Quality Assurance plan require that activities affecting quality be prescribed by documented procedures and be accomplished in accordance with these procedures.
Procedure SMAP-8, Issue 6,
" Plant Operations Shift Turnover," requires turnover of vital area and critical valve keys.
Contrary to the above, on September 9, 1986, at 8:00 a.m., the shift turnover was accomplished without the vital area and critical valve keys being turned over to the oncoming shift supervisor.
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P-87001 Page 2-
' January 5,-1986
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This is
- a. Severity Level.IV violation.
(Supplement I.E.)
(267/8625-01)
(1) The reason for violation if admitted:
The violai,on ~ was due to personnel error. The off-going Shift Supervisor simply forcot to turn over vital area and critical valve keys to the on-coming Shift Supervisor.
(2) Corrective steps which have been taken and results achieved:
The two Shift Supervisors received verbal reprimands from the Superintendent of Operations. The issue was discussed in detail with all Shift Supervisors at the Superintendent of Operations' staff meeting held on December 3, 1986.
The physical control of vital area and critical valve keys has been established at the Central Key Repository (CKR).
The CKR is co-located in the Security Badge Cube. Keys are issued and accounted for from the Security Badge ' Cube to-maintain more positive physical control.
Security Administrative Procedure Nine (SAP-9), " Security Key Issue", has been revised and is currently in effect.
SAP-9 provides for administrative control and documentation of the issuance of vital area and critical valve keys.
Superintendent of Operations Administrative Procedure Four (SOAP-4), " Plant Operations Shift Turnover Procedure", has been revised to reflect the change to SAP-9 and to coordinate the physical control of vital area and critical valve keys between security and personnel involved in shift.
turnover.
The revision to SAP-9 was issued effective November 12, 1986.
SOAP-4 has been revised and is now in the review / approval cycle.
(3) Corrective steps which will be taken to avoid further violations:
The actions described in 2) above will be sufficient to prevent further violations.
The revision to SOAP-4 to incorporate minor administrative changes will be issued by January 11, 1987.
.(4) Date when full compliance will be achieved:
Full compliance will be achieved upon the issuance of the revision to SOAP-4, 2.
Failure to Review Modification Control Procedures Technical Specifications, paragraph 7.4.b, requires that procedures and administrative policies of Appendix A
to Regulatory Guide 1.33, November 1972, which includes general
P-87001 Pag: 3 January 5 1986
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procedures for control of modification work, be reviewed - by the plant operations review committee (PORC).
Contrary to the
- above, procedures ENG-1
(" Control of Modifications and Documentation"). ED-100 ("CN Preparation and Document Control"),
and CWPM
(" Controlled Work Procedure Manual"), which all provide administrative policies of control of modification work, were not reviewed by PORC.
This is a Severity Level IV violation.
(Supplement I.D.)
(267/8625-02)
(1) The reason for violation if admitted:
The violation is not admitted.
In December, 1980, PSC explicitly considered whether or not Paragraph 7.4 of the Technical Specifications applied to Nuclear Engineering procedures. At that time, it was determined (as documented in Procedure G-2 " Fort St. Vrain Procedure Systems," Issue 3, Paragraph 4.3.3 - see Attachment 1) that the review by PORC of Level 2 engineering procedures, which provide detailed internal Nuclear Engineering Division requirements and are subordinate to Level 1 administrative procedures, was not necessary. The basis for this determination was that adherence to Level 1 procedures is mandatory (FSAR, Appendix B.5.2.2.2) and therefore, Nuclear Engineering procedures cannot conflict with Level 1 administrative procedures. This determination is also documented in the current issue of Nuclear Engineering Procedure ENG-6 "NED Procedure System," Section 4.3.3 b.
To implement the cited Technical Specification paragraph, level 1 administrative procedures Q-3 " Design Control System" (Attachment 2), and G-9 " Controlled Work Procedures" (Attachment 3), are reviewed by PORC.
In addition, each safety related Controlled Work Procedure (and the associated safety related Change Notice, which are prepared in accordance with the detailed Level 2 internal Nuclear Engineering procedures),
is individually reviewed and approved by PORC' prior to implementation.
The applicability statement of paragraph 7.4 of the Technical Specifications states that it is applicable to administrative procedures which govern plant operation.
Administrative Procedure G-1 " Definitions" further defines plant operating procedures as written instructions defining the normal methods, means, and limits of operations of a nuclear power plant.
Safety Guide 33, dated November 3, 1972 provides the following guidance:
(Excerpt from Arpendix A,
Paragraph I - Procedures for PerformingMaintenance)
P-87001
- z.
Paga 4-
.i January 5, 1986
" General,. procedures for the-control of maintenance, repair replacement, and modification work should be prepared before reactor operation is begun."
Thei above statement from Safety Guide 33 is the only explicit reference to modification procedures in this guidance document.
As described above, PSC considers that Procedures Q-3 and G-9 provides for the implementation of.
regulatory requirements related to modifications and design control. Engineering procedures, subordinate to Q-3 and G-9,.do.not govern plant operations and therefore do not require-PORC review in accordance with Specificaton Paragraph 7.4
- Technical (2) The corrective steps which have been taken and the results achieved:
Not applicable (3) Corrective. steps which will be _taken to avoid further violations:
Not applicable (4) The date when full compliance will be achieved:
Not applicable 3.
Failure to Sufficiently Document Design Verification:
Technical Specifications paragraph 7.4.a.1, requires that written procedures be established, implemented, and maintained for activities ' covered by Appendix A of Regulatory Guide 1.33, November 1972, which includes general procedures for control of modification work.
Procedure ENG-1, " Control of Modifications and Documentation,"
requires that the person performing the design verification process document their effort in sufficient detail to provide a record of their work.
Contrary to the above, the person performing the design verification for change notice (CN) 1876 did not document the effort in sufficient detail to provide a record of the work in that the " Check List of Design Verification Questions for Design Review Method" was signed and dated on one date and was initialed and dated on three subsequent dates with no details provided as
-to _what was verified on each date, and no details were provided to record that the design sketches included in the CN package were checked by the design verifier.
This is a
Severity Level IV violation.
(Supplement I.D.)(267/8625-03)
P-87001 Page 5 January 5, 1986 (1) The reason for violation if admitted:
The violation is not admitted. As cited in the violation, engineering procedure ENG-1 (see Attachment 4) prescribes the independent design verification (IDV) process for change notices (CNs).
Procedure ED-100 (Attachment 5) further describes the process for preparation, assembly, review and approval of change notice packages.
Paragraph 4.5.2 of ED-100 states that the coordination sheet of a CN documents the Coordinator, interface reviewers, page schedule, new pages, substituted pages and NED approval. This CN coordination sheet serves as a cover page for the three design packages I
comprising a CN.
The coordination sheet provides two things: 1)anindex identifying all the Change Notice pages and their issue that comprise the change notice package; and, 2) a record of the person /date performing the various duties, i.e.,
- IDV, interface review, and approval of the CN package. The signatures indicate they have reviewed / approved the package as required to fulfill their specific responsibilities as defined in the engineering procedures.
The signature by the person doing the IDV on both the IDV checklist page and on the coordination sheet for CN-1876 (see Attachment 6) explicitly documents that the independent design verifier has performed functions on all appropriate pages of the CN package, not just on the checklist pages.
In particular, each of the four design sketches in CN-1876 were assigned a CN page number, and each of those CN page numbers was explicitly included on the CN coordination sheets signed by the independent design verifier.
To provide a signature on each individual sketch would then place doubt in ones mind that any of the other pages of the CN package were reviewed.
PSC believes that we have performed all the requirements in the preparation and review of the subject Change Notice in accordance with procedures and regulations. However, the NRC has identified what PSC would consider to be a weakness, in that several IDVs/ reviews / approvals have been performed without detailed documentation identifying why the several signatures were required.
Because we believe this may be a weakness, the following actions have been initiated:
An instructional memo was distributed on December 9, 1986 (NDG-86-1647) to all individuals involved in
.he CN preparation process and all individuals who may perforin an IDV. The instructional memo reiterates the responsibilities of the individual performing the IDV and the documentation requirements.
- P-87001-Pag 2.6 January 5, 1986 In addition, our engineering procedures will be changed and a form created so'that more explicit documentation will be required when the CN package receives more than one set of.
'IDV/ review / approval. This permanent procedure change -will be implemented by January 19, 1987.
(2) The corrective steps which have been taken and the results achieved:
Although PSC does not' consider a violation to have occurred, the Nuclear Engineering Division has issued an instructional memo (NDG-86-1647). reiterating the design ~ verification process and procedural requirements.
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(3) Corrective steps which will be taken to avoid further violations:
Not applicable (4) The date when full compliance will be achieved:
Not applicable 4.
Failure to Periodically Test Flow Orifice Valve Limit::
Criterion XI to 10 CFR Part 50, Appendix B, requires that testing required to demonstrate that structures, systems, and components will perform satisfactorily in service be performed in accordance with written test procedures which incorporate the requirements contained in applicable design documents.
The licensee's approved quality assurance program states that testing activities are conducted on a continuing basis to satisfy Technical Specification requirements and, as appropriate, following modifications when specified by. the Change Notice package.
The safety evaluation for Change Notice (CN) 1876 states that the limit trip feature in the close position for the flow orifice valves is ~ required to be tested on a periodic basis to verify the design function of preventing an orifice valve from reaching the full closed mechanical'stop.
Contrary to the above, following modifications to the flow orifice valves control circuits, periodic testing of the flow orifice valve limit trip feature was not being accomplished as prescribed by the instructions contained in the safety evaluation for CN 1876.
This is a Severity Level IV violation.
(Supplement I.D.)
(267/8625-04)
(1) The reason for violation if admitted:
P-87001 Pag 2 7 January 5, 1986 The need to perform periodic testing of the new limit switches installed by CN-1876 was initially identified by the Fort St. Vrain Work Review Coninittee.
Physical manipulation of the orifice valves to accomplish limit switch testing is controlled by the Fort St. Vrain Technical Specifications, Limiting Conditions of Operation (LCO) 4.1.9.
It is not desirable to actually stroke the orifice valves to the required closure for limit switch actuation during plant operations or during outages since this would involve entering a degraded mode of LC0 4.1.9.
Recognizing
- this, it was intended to test and calibrate the Control Rod l
Drive protective features when the associated assemblies were removed from the core for preventive or corrective maintenance. The use of a simulated input for the above surveillance to test the limit switch feature was not considered.
(2) Corrective steps which have been taken and results achieved:
Surveillance Test SR-RE-160-SA, " Functional Test of CRDA Orifice Limit Switches", has been written.
This new l
procedure will test the Orifice Valve limit trip feature by simulating valve closure (as is done in the calibration of the limit switch) and verifying that the valve will not move further in the closed direction.
It should be noted that the limit switches have been tested on two previous occasions. The switches were first tested upon installation in August, 1985. They were tested again in March, 1986 when their reset functions were recalibrated.
SR-RE-160-SA has been written and routed for administrative approvals.
The new test was approved by the Plant Operations Review Committee (PORC) on December 23, 1986.
(3) Corrective steps which will be taken to avoid further violations:
The procedure has been approved and was issued on December 24, 1986,, and will then be scheduled for completion on a semi-annual basis.
f (4) Date when full compliance will be achieved:
Full compliance was achieved by December 24, 1986 with the issuance of SR-RE-160-SA.
5.
Failure to Inplement Corrective Action Program Criterion XVI to 10 CFR Part 50, Appendix B, requires that measures be established to assure that conditions adverse to
- quality, such as
- failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified ar.d corrected.
r:
P-87001 Pag 8 January 5, 1986 Contrary to the above, the installation instructions for CN 1876 noted that numerous discrepancies were identified with the existing vendor design documents and the actual field installation, but no corrective action was initiated to review other areas of the vendor documents not affected by the change notice) for similar discrepancies.
This is a
Severity Level IV violation.
(Supplement I.D.)(267/8625-05)
(1) The reason for the violation if admitted.:
The violation is not admitted. CN-1876 identified several discrepancies in a common point connection.
These discrepancies were between the way in which the field wiring was actually connected and the way it was represented on our vendor wiring lists.
At the time the design modification was being prepared,. the discrepancies were identified because the designer was determining which changes had to be made to implement his new design.
When he recognized differences between the field and documentation he performed addition field verification work to determine if other discrepancies existed.
This field verification was performed on the relay panel in the Auxiliary Electric Room, and the applicable portion of the Control Broad involved with this design change. Some areas were not accessible because wire bundles would have to be breken to trace wiring and the hazard of inadvertently tripping operating circuits.
Because of this inability to verify all the circuits, the statement requesting that construction personnel verify connections before removing any terminations or removing equipment was placed in the Change Notice. As a result, one additional wiring deficiency was identified. The CN was modified by a Deviation Request to document the deficiency.
It should be noted that the wiring deficiencies being discussed were common point termination differences and not differences with the circuit design and therefore the electrical system is installed and operating per the original design.
(2) The corrective steps which have been taken and the results achieved:
Not applicable.
(3) Corrective -teps which will be taken to avoid further violations:
Not applicable.
(4) The date when full compliance will be achieved:
Not applicable,
P-87001-.
Pags 9--
January 5, 1986 Should you have any questions concerning this response, please
. contact Mr.
M.
H.. Holmes at (303) -480-6960 for additional information.
Very truly yours, N. L. iMy&M& MAO H. L. Brey, Manager Nuclear Licensing and Fuels Division HLB /DCG:jmt Attachments
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