IR 05000285/1988032
| ML20206A416 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 11/03/1988 |
| From: | Harrell P, Reis T, Westerman T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20206A406 | List: |
| References | |
| 50-285-88-32, GL-88-03, GL-88-3, IEB-88-005, IEB-88-5, NUDOCS 8811150111 | |
| Download: ML20206A416 (27) | |
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APPENDIX B U. S. NUCLEAR REGULATORY COMMISSION
REGION IV
NRC Inspection Report: 50-285/88-32 License:
DPR-40 Docket: 50-285
Licensee: Omaha Public Power District 1623 Harney Street Omaha, Nebraska 68102 Facility Name:
Fort Calhoun Station (FCS)
Inspection At:
Fort Calhoun Station, Blair, ilebraska Inspection Conducted: September 1-30, 1988 Inspector:
Q EL(off.g
. /0 - /2 -86 PfrHarrell, Senior Resident Reactor Date
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anspector
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ns C Reis, Residsnt Reactor Inspector Date Approved:
7. N /d d2Z. i m
//- E-6 S T. F. Westerman, Chief Project Date
Section B Civision of Reae. tor Projects
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Inspection Sununary Inspection Conducted September 1-30, 1988 (Report 50-285/88-32)
Areas inspected: Routine, unannounced inspection including followup on previously identified items, operational safely verification, plant tours, monthly maintenance cbservations, monthly surveillance observations, security observations, radiological protection observations, in-office review of periodic and special reports, review of Generic Letter 88-03, review of NRC Bulletin 88-05, review of information on containment temperatures, and preparation for refueling.
Results: During this inspection period, the licensee continued to demonstrate a lack of attention in maintaining fire barriers in a functional status.
The NRC inspector identified a fire door that would not latch due to modifications performed during security door alterations.
The licensee did not issue a report to the NRC on the nonfunctional fire barrier as required by the Technical Specifications. The licensee experienced similar problems in the area of reporting nonfunctional fire barriers in that Violation 285/8621-02 was issued on this subject in July 1986. A violation was issued in paragraph 5.a of this report for failure to issue a report to the NRC.
It was also noted during this inspection period, that the licensee did not perfonn an hourly fire watch patrol for all nonfunctional fire barriers as required by the Technical Specifications. The itcensee also experienced problems of this nature in that Violation 285/8603-01 was issued on this subject in February 1986. A violation was issued in paragraph 5.b of this report for failure to perform hourly firn watch duties.
In both cases discussed above, it appeared that the licensee did not adequately implement the corrective actions in response to violations issued by the hRC. The report of these violations is an indication that the licensee needs to concentrate on effectively implementing their corrective action program.
In paragraph 5.d of this inspection report, a violation was issued for modification of the plant without proper work instructions. The modification involved renoval of a pipe snubMr installed in the auxiliary steam system.
Although this is the first time this problem has been idenLified, the licensee needs to take actions to ensure that all nodifications made to the plant are adequately addressed by appropriate procedures.
The plant operated at power centinuously for 477 days,13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />, and 15 minutes.
Tnis power run established a record for all nuclear plants.
In paragraph 9.b. of this inspection report, an open item was issued related to the posting of radiological hot spots in the plant. The open item was related to the failure of the licensee to ensure the hot !. pot stickers were visible from all possible vantage points.
The licensee inmediately corrected the concern identified by the NRC inspector; however, a change to the radiological protection manual should be made to include specific instructions fer hot spot posting.
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During this inspection period, the licensee shut down the plant to comence a refueling outage. A r.'"iew of the activities associated with the refueling
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outage indicates that the licensee was properly prepared by issuonce of the
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appropriate procedures.
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DETAILS 1.
Persons Contacted OPPD
- X. Morris, Division Manager, Nuclear Operations
- W. Gates, Plant Manager J. Bobba, Supervisor, Radiation Prutection
- J. Fisictro, Manager, Nuclear Licensir.g and Industry Affairs
- S. Gambhir, Division Manager, Production Engineering
- L. Guach m, Plant Licensing Engineer
- R. Jaworski, Manager, Technical Services Station Engineering J. Kecy, Supervisor, System Engineering
- A. Richard, Manager Quality Assurance and Quality Control C. Simons, Plant Licensing Engineer
- J. Smith, Manager, Security Services
- S. Willrett, Manager, Administrative Services
- Denotes attendance at the monthly exit interview.
The NRC inspectors also contacted other plant personnel, including operators, technicians, and administrative personnel.
2.
Plant Statu_s_
On September 27, 1988, the licensee comenced reducing power from 30 percent to start its eleventh refueling outage.
The plant entered Mode 2 at 5:35 p.m. on September 27, 1988, and Mode S at 9:30 p.m. on October 1, 1988. The plant had been in continuous power operation (Mode 1) for 477 days,13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />, and 15 minutes.
The refueling outage is currently scheduled to last 77 days, with the plant returning to 100 percent power by December 31, 1988.
In addition to refueling, the major work scheduled for the outage includes:
renmal anu rebuilding of two reactor coolant pump motors; overhaul of both ercrgency diesel generators; eddy current testing and sludge lancing of both steam generators; inspection of components in the instrument air system; installation of control room design modifications; replacement of the control room heating, ventilating; and air conditioning system; and work on the, main turbine.
3.
Followup on Previously Identified Items (92701)
a.
(Closed) Open Item (285/8803-02): Valve CH-198, a Check Valve, is Being Used as a Containment Isolation Valve - This item involved the installation of a check valve in the charging line that penetrates containment. The valve has been installed since original
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r construction. The check valve is used as a containment isolatu n valve which appears to be contrary to the requirements of Criterion 55 of Appendix A to 10 CFR Part 50.
j On January 10, 1986, the NRC's Office of Nuclear Reactor Regulation (NRR) issued a safet/ evaluation report (SER) that addressed the use of Valve CH-198 as a containment isolation valve.
i The SER was issued in response to a request by the licensee to exempt
Valve CH-198 from the local leak rate testing required by Ap u dix J l
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to 10 CFR Part 50. NRR concluded that testing of Valve CH-198 was not required.
The SER stated that the justification for not testing this valve is that the pressure (2100 psig) seen by the valve in the direction of
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accident pressure (60 psig) greater than the maximum containment flow toward containment is f
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All of the charging pumps remain
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operational or are automatically started and Valve CH-198 remains open upon receipt of a safety-injection actuation signal (SIAS).
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Thus, the chargin;r pump flow provides a seal barrier against escape of the containment atmosphere. Maintaining this barrier during a
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loss-of-coolant accident is ensured since-upon recaipt of an SIAS, i
the charging pumps are automatically aligned to the boric acid storage tanks. The volume held by these tanks provides a source of
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supply to the charging pumps for approximately 80 minutes and, as demonstrated in Section 14.16 of the Updated Safety Analysis
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Report (USAR). the containinent pressure would be reduced back to near atmospheric levels (approximately 2 psig) within 50 minutes.
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after the tanks are empty, there will exist a 14-foot head of water on the suction side of the charging pumps.
This head of water will
exert a pressure of approximately 6 psig to provide a seal against air leakage for the remainder of the accident.
Based on the review performed by NRR, it appeers that the use of a f
Valve CH-198 as a containment isolation valve is acceptable, j
b.
(0 pen) Deviation (285/8811-02):
Erection of Scaffolding Without the
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Performance of a Safety Review - This item concerns a deviation from (
a comitment made to the hRC.
In a letter, dated February 24, 1988, i
the licensee submitted a response to Violatio.s 285/8724-05.
In the
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response, the licensee stated that a review would be perfonned prior to the erection of scaffolding in safety-related areas for
maintenance or any other activities.
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In deviation from the above, on March 15, 1988, the NRC noted that scaffolding was erected for painting of the errergency feedwater storage tank, which is a safety-related area, and no prior review had I
been performed. Subsequent to notification of this deviation by the
NRC inspectors, the licensee renoved the scaffolding without incident.
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Prior to the issuunce of Violation 285/3724-05 and this deviation, a
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procedure for the control of scaffolding at the FCS did not exist.
Scaffolding was erected on an as-needed basis by the respective crafts.
To preclude recurrence of scaffolding being erected without a safety evaluation when required, the licensee has designated scaffolding as a temporary mechanical jumper to be controlled by Procedure 50-0-25
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"Temporary Modification Control."
A procedure change, issued August 5,1988, required that a cognizant engineer perform a safet;' evaluation pursuant to 10 CFR Part 50.59 for scaffolding erected tver or near CQE (safety-related) or fire protection equipment.
However, the NRC inspector noted that the procedure revision lef t an
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unacceptable provision in the process.
Paragraph 2.2.10 of
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Revision 22 to Procedure 50-C-25 states that the provisions of the chapter are not applicable for scaffolding erected where there is no CQE or fire protection equipmen! under or near the scaffolding.
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NRC inspectors felt that the paragraph as written could allow
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unqualified personnel to make the decision as to what equipment is CQE or fire protection.
This procedural deficiency was discussed with the licensee engineer responsible for the revision. The engineer agreed that the revision, as written, left the potential for
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erection of scaffolding in an unreviewed manner. The engineer agreed to further revise the procedure.
This item remains open pending
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review of the forthcoming revision.
c.
(Closed) Severity Level V Violation (285/8811-03):
Failure to make a 4-Hour Report - This item was related to the licensee's failure to make a 4-hour report in accordance with 10 CFR Part 50.72(b)(2). The report should have been made when the licensee inadvertently started Emergency Diesel Generator (EDG) 2 during the testing of EDG 1.
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The licensee issued Licensee Event Report (LER)88-007 to describe the events related to this violation.
The licensee described in LER 88-007, the corrective actions that are planned to ensure
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compliance with Part 50.72(b)(2).
The specifics identified in this violation will be reviewed during closeout of LER 88-007; therefore, this violation is considered closed, d.
(Closed) Open item (285/8811-09): Clarification the Role of (
Knowledgeable Person in Reviewing Procedures Contained in the i
Operating Rinual - This open item originated from a concern addressed as an allegation that operations personnel were assigned procedures
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in the operating manual for review to determine such criteria as where quality control (QC), quality assurance (QA), and radiological holdpoints should be included.
The allegation was that operations
pcrsonnel are not qualified to perform this function.
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Paragraph 5.2.15 of ANSI N18.7-1976 requires that plant procedures be
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reviewed by an individual knowledgeable in the area affected by the
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procedure no less frequently than every 2 years to determine if
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changes are necessary or desirable. ANSI 18.7-1976 does not provide
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any guidance or definition as to what constitutes a knowledgeable
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individual. The licensee's procedure governing the review program,
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Procedure 50-G-36, "Operating Manuals Review Documentation," also i
does not provide guidance as to what constitutes i knowledgeable
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individual.
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To address the concern, the licensee has interp'eted a knowledgeable
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individual as a person who would use the operating manual in the
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Furthemore, t le licensee has
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perfomance of his job duties.
i required supervision to designate who the knowledgeable individual to l
perfom specific reviews will be.
The licensee issued a change to Procedure 50-G-36 on July 25, 1988, which serves the following functions:
Defines knowledgeable individual
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Designates the cognizant supervisor for the individual volumes
J of the operating manual Provides instructions to assigned reviewers to seek guidance
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cannot resolve The NRC inspectors reviewed the procedure change. Based on the review, it appeared that the licensee's policy meets the requirements of ANSI N18.7-1976 and tne licensee has clarified the assignment of
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i knowledgeable individuals as document reviewers, i
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(0 pen) Severity Level IV Violation (285/8729-04):
Failure to issue l
and Perfom Required Surveillance Testing Corres 30nding to a Technical Specification (IS) Amendment - On Octo>er 30, 1987
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Amendment 110 of the TS became effective and established the i
surveillance requirements for the core exit and heated junction l
thermocouple instrumentation systems. The lice 1see failed to issue i
and perform the required surveillance tests. On Novect>er 10, 1987, the NRC inspectors made the licensee aware of this delinquency.
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The licensee admitted to this violation in response to the Notice of
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l Violation in NRC Inspection Report 50-285/87-29. The licensee stated that the root cause for the violation was that the appropriate f
personnel were not aware of the issuance of the TS amendment due to
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an error in correspondence disi:ribution.
l The irrediate corrective action taken by the licensee was to issue
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and perfom the procedures and procedure changes necessary to comply
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with TS Amendment 110. The licensee issued the following procedures and performed testing to verify instrumentation operability on November 24, 1987:
ST-CET-1, Revision 0, "Heated Junction Thermocouple Channel
Check" ST-HJTC-1, Revision 0, "heated Junction Thermocouple Channel
Check" ST-SM-1, Revision 5, "Subcooled Margin Mater Check"
OP-1, Revision 32 "Master Checklist for Start-up or T ip
Recovery" The NRC inspectors reviewed these procedures and procedure changes in coordination with TS Amendment 110 and found them to meet the requirements as set forth in the TS.
The licensee further committed to resolve the generic problem of correspondence control and has taken several actions concerning the administrative contiol of TS amendments to date.
On December 31, 1987, the licensee issued Revision 1 to Procedure G-2, "Integrated Regulatory Requirements Log," of the Nuclear Operatiens Division Policy and Procedure Manual, which required the following actions be taken upon receipt of an amendment to the TS:
A Regulatory Requirement Document (RRD) shall be issued to the
Supervisor-Technir:al to ensure reviews and procedure revisions are completed within the implementaticn period.
An RRD shall be issued to the training department for personnel
training.
An infonnational copy shall be provided to the
Supervisor-Operations utilizing an ExD to acknowledge receipt.
The NRC inspectors reviewed this procedure change and found that it provided an appropriate avenue to enture the cognizant personnel are aware of TS amendtrents so that procedural recuirements may be implemented.
On August 19, 1988, the licensee issued Revision 3 to Procedure G-2 which reassigned the responsibilities of the Supervisor-Technical to the Assistant Plant Manager. The fomer position has been abolished.
The generic problem of correspondence distribution was addressed by licensee internal Memorandum LIC-80-261.
In this discussion, the licensee contended that current method of correspondence distribution
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is basically sound. The error which generated this violation was found to be an isolated personnel error and not a programatic issue.
In the analysis, however, the licensee did find ways to improve the program of corresponder.ce control and distribution. The improvement was in the area of developing formal procedures for correspondence distribution.
A proceduralized program has been established for the distribution of TS Amendments, which is the basis for this violation. However, the licensee stated that formalized procedures for the control of all external correspondence will be issued in the near future. This item remains open pending a review of the licensee's procedures for control of external correspondence.
Operational Safety Verification (71707)
The NRC inspectors conducted reviews and observations of selected activities to verify that facility operations were perfonrad in conformance with the requirements established under 10 CFR, the licensee's administrative procedures, and the TS. The NRC inspectors made several control room observations to verify the following:
Droper shift staffing
Operator adherence to approved procedures and TS requirements
Operability of reactor protective system and engineered safeguards
equipment Logs, records, recorder traces, annunciators, panel indications, and
switch positions complied with the appropriate requirements Proper return to service of components
Maintenance orders (MO) initiated for equipnwnt in need of
maintenance Appropriate conduct of <:ontrol room and other licensed operators
Management pertonnel toured the control room on a regular basis
On September 2, 1988, the licensee was notified by the contractor perfonning the design basis reconstitution that the plant was not within the established design basis. The contractor noted that the licensee did not have sufficient fuel oil stored on site to neet the requirement established in the Basis Section of TS 2.7.
The Basis Section states, in part, that the 16,000 gallons in the fuel oil storage tank, in addition to the day tanks, will provide diesel operation under the required loading conditions for a minimum period of 7 days should only one diesel be in operation.
The Basis Section also states that it is considered incredible
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not to be able to secure fusi oil from one of several sources in the vicinity of Omaha in less than 3 days under the worst weather conditions.
The contractor perfomed calculations and determined that EDG 1 would require a total of 18,394 gallons and EDG 2 a total of 18,896 gallons of fuel oil to meet the 7-day requirement. The total amount of fuel oil available on site for either EDG is 18,850 gallons. As a result, the supply available for EDG 2 is 46 gallons short of a 7-day supply.
The licensee is in compliance with the requirements of TS 2.7 in that the amount of fuel oil requirea by the TS is 16,000 gallons available in the fuel oil storage tank and 550 gallons available in the EDG base tank.
This amount of fuel oil providet, a supply cf approximately 6 days. The licensee has maintained that this amount of fuel oil onsite is all that is required by TS 2.7.
The licensee performed a review of the situation and determined that both EDGs were considered operable. The determination was based on the statement in the Basis Section of TS 2.7 that it is considered incredible not to be able to secure fuel oil within 3 days under the worst of weather conditions.
The NRC inspectors discussed the situation described above with NRC Region IV and NRR personnel. The discussion established that since the licensee met the requirements of TS 2.7, the EDGs were not considered inoperable nor was the licensee required to initiate a justification for continued operation.
In response to this situation, the licensee submitted a request for amendment of TS 2.7 on September 9, 1988. The letter requests that the Basis Section of the TS be amended to change the statement that a 7-day supply is available to a statement that a 6-day supply of fuel oil is available on site. NRR is currently reviewing the licensee's submission.
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No violations or deviations were identified.
5.
Plant Tours (71707)
The NRC inspectors conducted plant tours at various times to assess plant and equipment conditions. The following items were observed during the tours:
General plant conditions, including operability of standby equipment,
were satisfactory.
Equipment was being maintained in proper condition, without fluid
leaks and excessive vibration.
Plant housekeeping and cleanliness practices were observed, including
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Performance of work activities was in accordance with appreved
procedures.
Portable gas cylinders were properly stored to prevent possible
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missile hazards.
Tag out of equipment was performed properly.
- Management personnel toured the operating spaces on a regular basis.
- During tours of the plant, the NRC inspectors ncted the items listed below:
a.
During a plant tour on September 1,1988, the NRC inspectors noted that Fire Door 1007-16 would not latch.
The failure of the door to latch rendered the door inoperable and thus rendered the fire barrier associated with the fire door inoperable.
The fire barrier protects safety-related equipment.
Upon notification to the licensee of the nonfunctional fire door, the licensee performed a review to determine the reason for the nonfunctionality.
The licensee determined that the fire door became nonfunctional on July 15, 1988. The licensee established that on this date, the electromechanical door latch was elr.ctrically disconnected when modifications were made to the site security system.
Electrically disconnecting the latch prevented the door from being held closed. The modification of the latch was documented on MO 883100.
TS 2.19(7) states, in part, that all p'netration fire barriers prctecting safety-related areas shall De functional (intact).
If nonfunctional, restore the penetration to a fcnctional status within 7 days, or prepare and submit a report to the Comission within an additional 30 days.
Contrary to the above, Fire Door 1007-16 was not ispaired within 7 days and a report was not submitted to the Commission within the next 30 days. This is an apparent violation.
(235/8832-01)
During the inspection period of July 1-31, 1986, the NRC inspectors noted the same violation of TS 2.19(7) as described above.
The details of the previously identified violation art described in NRC Inspection Report 50-285/86-21.
It appears that the corrective actions were not properly implemented by the licensee to ensure compliance with the TS requirement of reporting nonfunctional fire barriers.
On September 1, 1988, the licensee issued M0 883916 to provide instructions for repairing the latch.
The licensee completed repair of the latch and returned the fire door to a functional status on September 2,198, _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ - - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - __ ______ _____
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The licensee reviewed the problem discussed above to..termine why the nonfunctional fire door was identified on July 15, 1988, and immediate action was not taken to repair the door. The licensee determined that MO 883100 had beca classified by the individual initiating the M0 as a structural M0. At the time the MO was issued, the fire protection engineer did not routinely review M0s classified as structural.
To ensure that nonfunctional fire doors and barriers are promptly repaired, the licensee instructed the fire protection engineer to routinely review M0s that are classified as structural, b.
During a tour of the contml room on September 29, 1988, the NRC inspectors reviewed the H e iy Fire Watch Patrol Log, Form FC-1006, maintained by security pet..:nel.
The log is maintained by security since these personnel perfom the Cuties of tLe hourly fire watch.
The log was established by the licensee to list all nonfunctional fire barriers and to provide a signature record that each nonfunctional barrier was checked hourly.
During a review of Fom FC-1006, the NRC inspectors noted that security personnel had not checked all nonfunctional fire barriers during the hours of 10 a.m. to 12 p.m.
The log noted that all of the 32 inoperabia fire barriers were not checked because the security computer was inoperable.
TS 2.19(7) states, in part, that all penetration fire barriers protecting safety-related areas shall be functional (intact). With a penetration barrier nonfunctional, verify the opert.5111ty of fire detet. tors on at least one side of the penetration and establish sn hourly fire watch patrol.
Contrary to the above, the licensee's fire watch patrol was r-ot provided during the hours of 10 a.m. to 12 p.m. for all nonfunctional fire barriers on September 29, 1988.
This is an apparent violation.
(285/8832-02)
During a review of this problem, the NRC inspectors noted that the Supervisor of Security issued a memorandum on September 10, 1988, to alert all security personnel of the TS requirecent to check all nonfunctional fire berriers hourly. The memorandum was issued based on a concern identified by NRC inspectors during a tour of the auxiliary building on September 9,1988. The remorandum stated that security personnel should ensure that the shift supervisor was notified whenever a fire barrier could not be checked.
It did not appear that these actions were completed when the fire barriers were not checked on Septerber 29, 1988.
In an NRC inspection perfomed during the period of Feoruary 1-28, 1986, a violation of TS 2.19(7) was identified with respect to tne failuri of the licensee to perform hourly checks of nonfunctional
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It appears that the corrective actions were not p operly implemented by the licensec to ensure compliance with the TS n quirement for checking nonfunctional fire barriers hourly.
I' response to this identified problem, the licensee issued a security bulletin, dated October 1,1988, to notify security personnel that the fire door patrol would not be removed from his/her i
duties for any reason. The bulletin also stated the fire door patrol
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will not be assigned any additional duties and will not be a member I
of the response force. The actions are intended by the licensee to ensure all nonfunctional fire barriers are checked hourly, c.
On September 6, 1988, the NRC inspectors noted that a nitrogen gas cylinder located in the auxiliary building was not secured. The cylinder was pressurized and in service at the time of discovery.
In NRC Inspection Report 50-285/87-27, a violation was issued for
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failure to control the storage of compressed gas cylinders. Since the violation was issued, this has been the only time noted by the NRC inspectors where inad uate control of.:ylinders has been identified.
For this rea this one problem was not considered a programmatic breakdown in the requirements to properly control gas
cylind_rs.
l To ensure compliance is maintained in the future, the licensee i
implenented a formal requirenent where the auxiliary building watchstander routinely verifies proper storage of gas cylinders.
d.
On September 9,1988, the NRC inspectors nated that a support for an auxiliary steam line was disconnected. That is, a partial snubber
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was found to be attached to the auxiliary steam line, but the support was not anchored to any foundation. A review of the engineered drawing for this support, designated as AXS-10, found that it was a
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i seismically designed support. The NRC inspectors were conearned that the failure of this line could impinge on safety-related equipment
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in the area, i
A review of past work instructions made it apparent that the support
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was removed without work instructions a ad without an engineering analysis having been performed. On October 30, 1987 MO 875029 was initiated identifying AXS-10 as having a loose ceiling anchor bolt.
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The M0 provided instructions to retorque the loose bolt to I
90 foot-pounds and to detemine the integrity of the anchor.
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work acconplished section of the MO, it was verified that the bolt
was torqued per instructions.
It further stated that "removed both
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ceiling bolts and wired hanger off to handrails so that inspection could be made." There was no documentation of removal of the i
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snubber. The work was docurr.ented as cortpleted by the craf t supervisor on January 11, 1988.
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On September 9, 1988, the NRC inspectors notified the licensee of the apparent deficient condition.
On September 9, 1988, a safety i
evaluation was generated by engineering and concluded that there were
no safety implications resulting from the removal of tne pipe support t
provided the line was not energized (i.e., not pressurized with
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auxiliary steam). Operations personnel prevented the line from being
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energized by danger tagging the steam supply valve to this line. The i
NRC Inspectors reviewed the engineering analysis and found that it L
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adequately addressed the concerns.
On September 29, 1988, the NRC inspectors interview 6d the lead
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craftsman who signed MO 875029 as complete. He recalled that he did not leave the support in the as-designed position. Rather, he left it tied off to the handrail so that engineering could inspect the anchor area. When asked why the MO,vas signed off as complete when the work was never performed, the lead craftsman statt.d that, based on his recollection, it was signed off so the papemork could clear and engineering could track the problem via an alternate document.
During the interview with the lead craftsman, another craftsman interjected that it was he who removed the snubber from its tied off position. He stated that he was verbally directed to do so by his supervision.
An engineering evaluation request EEAR-FC-88-007, was initiated on January 11, 1988, and evaluated on February 22, 1988. The evaluation recomended that AXS-10 snubber be redesigned and relocated via Modification Request MR-FC-88-007. MR-FC-88-007 was never issued and it was recomended that it be voided by Production Engineer Division Memorandum PED-FC-83-622 dated September 21, 1988.
This memorandum further concluded that the removal of Seismic Support AXS-10 would not reduce the margin of safety.
Therefore, the auxiliary steam line in the vicinity of AXS-10 was an unanalyzed condition from at least Octobor 30. 1987, until Septerrber 21, 1988.
TS 5.8.1 requires wrttten procedures and administrative policies to be established, implemented, and maintained that meet or exceed the minimum requirements of Section 5.1 of ANSI 18.7-1972.
Paragraph 5.1.6 of ANSI 18.7-1972 states, in part, that maintenance or modifications that may affect functioning of safety-related structures, systems, or components shall be perfonted in a manner to assure quality at least equivalent to that specified in applicable codes, bases, standards, and design requirements.
it further states that maintenance that can affect the perfornnte of safety-related equipmert shall be properly preplanned and perfonted in accordance with written procedures, documented instructions, or drawing appmpriate to the circumstance _ _ _ _____-____ -_ - - _ _ - _ _ _ _ _ _ _ _ _ _ _
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Contrary to the above, the licensee remved Seismic-Support AXS-10 without regard to original design requirements and performed this rodification without the use of written instructions.
This is an apparent violation.
(285/8832-03)
Based on the following considerations, the NRC inspectors concurred with the engineering evaluation that the snubber was not needed:
Review and verification of Memorandum PED-FC-88-007 which stated
in part that deadweight, thennal, and high energy line break loads are not appitcable to the evaluation of the snubber.
Yerification that the auxiliary steam line in question was not
Seismic Class I as defined in the Updated tifety Analysis Report. Appendix F of the USAR.
Discussions with the Manager of Mechanical Engineering,
Production Engineerirg Division, concerning his calculations that the line in question has no seismic interaction'.
6.
Monthly Maintenance Observations (62703)
The NRC inspectors reviewed and/or cbserved selected station maintenance activities on safety-related r.ysteia and components to verify the maintenance was conducted in accordance with approved procedures, regulatory requirements, and the TS. The following items were considered during the reviews and/or observations:
The TS limiting conditions for operation were met while systems or
components were removed irom service.
Approvals were obtained prior to initiating the work.
- Activities were accomplished using approved M0s and were inspected,
as applicable.
Functional testing and/or calibrations were perfonned prior to
returning components or systems to service.
QC records were maintained.
- Activities were accomplished by qualified personnel.
- Parts and caterials used were properly certified.
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Radiological and fire prevention controls were implemented.
- The hRC inspectors reviewed and/or observed the following raintenance activities:
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_ _ _ _ _ _ _ _
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L Installation of fire drtoctors in Panels Al-65A, AI-658, AI-66A, and
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AI-66B (M0 883933)
l InspectionofYalveHCV-2977(H0882479)
ReplacementofFireDoor989-12(MO8758?2)
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Replacement of hardware on Fire Door 989-12 (M0 875823)
Inspection of Valve HCV-484 (M0 880079)
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Inspection of Valve HCV-2967 (MO 882478)
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Painting of ?or.m 50 (M0882104)
Diesel generator mechanical modifications (MR-FC-87-38)
A discussion of each item is provided below:
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i a.
On September 2,1988, the !icensee found that the fire uetection
,
system did not conform with the design basis as stated in l
Section 9.11.2.2 of the USAR. The USAR stated that battery-powered
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smoke detectors have been installed in safety-related cabinets in the
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control room to provide early warning of smoke within the cabinets.
During a review of a concern identified by an NRC fire protectirn inspection team relating to smoke detectors in control room cabinets,
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i the licensee identified four safety-related cabinets that did not
have dt.tectors installed in them.
Upon identification of the problero, the licerisee issued MO 883933 to install the detectors.
The detectors were installed on Septettber 2 l
1988.
The NRC inspectors verified that the installations ware l
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l completed and that the detectors were tested af ter installation. A review of the MO indicated that the results of the werk had been properly documented and that the material used had been entered on
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the M0. No problems were noted,
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b.
On Septerrber 13, 1988 the licensee perfortred an inspection of the internals of Valve HCV-2977 in accordance with Procedure MP-CV0P-INSP. "Inspection of Internals of Cylinder and
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Piston-Operated Yalve Actuators." Procedure MP-CV0P-lhSP was
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attached to NO 88?479.
The inspection was performed to determine if l
the water intrusion into the instrument air systen had a detrinental
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affect on the valve. A fiber-optic sccpe was used to look for water i
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and/or corrosion. 1he licensee found no evidence of water and/or
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corrosion in the valve operator.
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The NRC inspectors observed that the licensee properly completed Procedure MP-CV0P-INSP by following each step as written and performci each step in order. The MO was properly completed to document the work performed.
The NRC inspectors used the fiber-optic scope to view the internals of the valve operator.
The NRC inspectors found that there was no water or corrosion present.
The internals of the valve operator were in very good condition.
The motite force for Valve HCV-2977 is instrument air.
The instrument air system was contaminated with water from the fire water system in July 1987. The inspection of the valve operator for Valve HCV-2977 was part of a program to inspect all instrunent air actuated valve operators suspected to have been affected by the water intrusion.
During this inspection period, other valve operators we 'e inspected for the presence of water and/or corrosion. The results of the other inspections are discussed in paragraphs d and e below, c.
On September 8,1988, the NRC inspectors observed licensee personnel remove Fire Door 989-12 and replace it with a new one. The replacement of the door was performed in accordance with MO 875822.
New door hardware (i.e., hinges, latch, and closing mechanism) was also installed. The hardware was installed in accordance with MO 875823.
l The NRC inspectors verified that the shift supervisor had authorized replacement of the door and that an hnurly fire patrol had been assigned while the door was nonfunctional.
The NRC inspectors also verified that the personnel installing the door properly followed the procedure as written and used approved materials.
No prcblems were noted.
d.
On September 27, 1938, the NRC inspectors observed the disassembly of the operator for Valve HCV-484 The valve operator was disassenbled in accordance with the instructions provided on MO 880079. The valve operator was disassemblet to inspect for the presence of water and/or corrosion.
The N?C inspectors noted that valve operator disassenbly was pert.
. d in accordance with the instructions, as written.
Verification that the valve had been properly isolated by the use of danger tags was perforr.cd.
The NRC inspectors also verified that the raintenance was performed by qualified individuals.
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In*,pection of the valve operator after disassembly indicated that no i
water was present. A slight amount of rust was preJent; however, the i
amount did not appear to affect the operation of the valve operator, f
No problem were noted during the reviews and observations rude by l
the NRC inspectors.
l P
e.
On September i4.1988, the NRC inspectors witnessed the inspection of the internals of the operator on Valve HCV-2967 The inspection was done under the authorization of MO 882478 which directed partial
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disassembly and inspection of the air operator per
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Procedure HP-CV0P-INSP. The NRC inspectors noted that the valve, the suction isolation valve for Containment Spray Pump SI-3B, was l
properly isolated tagged, and released by operations prior to
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starting the work.
The internals of the air operator were inspected with the use of a fiber-optic scope by a QC inspector, two machinists, a contract valve
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specialist, and the NRC inspectors. The internals of the piston-type i
actuator were genera'ly in good physical condition. However, the
personnel inspecting apeed to the existence of small patches of l
tightly adherent surface corrosion and droplets of moisture
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intermingled with the grea3e of the internals.
No other foreign
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material was noted.
The licensee stated that engineering would
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evaluate any impact the noted corrosion and water droplets would have on the operator. This item will be followed up during the closure of
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Open Item 285/8827-03, i
i f.
On Septerber 22, 1988, the NRC inspectors observed a portion of the i
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painting out of Room 59 in accordance with M0 862104 The NRC
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inscectors witnessed a health physics (HP) technician survey the hign-radiation area adjacent to the boronometer and counsel the I
painters on their task in the area. The craftsmen then employed
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extens.on handles on their paint rollers to increase their distance from the radiological source. The acttons were perceived by the NRC r
inspectors to be sound radiological work practices.
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g.
On September 22, 1988, the NRC inspectors were notified by licensee i
managenent that EDG 1 had been declared inoperable. The inoperability was based on the status of the diesel exhaust manifold which was inadequctely supported, from a seismic perspective, due to i
personnel error in perfonning a modification on the seismic.lupports f
for the exhaust piping. The personnel error stenred from failure to adhere to the written procedure by the craftsnen involved.
After physically inspecting the exhaust manifold and its supports, the NRC inspectors perceived the declaration of diesel inoperability to be conservative. Additional actions taken by the licensee included stopping the job to investigate the root cause of the procedural noncompliance and the retraining of the two contractor
~r crews perfonning the modification on the requirements for following l
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procedures exactly as written and to thoroughly read and understand procedures prior to embarking on the job. The root cause of the procedural noncompliance was found to be that the supervisor had not read the instructions provided in Modification Request MR-FC-87-38 thoroughly prior to starting work and therefore misinterproted a step when taken out of context of the complete procedure. Af ter declaring EDG 1 inoperable the operability of EDG 2 was proven as required by the TS.
Based on the review by the NRC inspectors, it appears that the licensee took th appropriate aci. ions in identifying, correcting, and reporting the procedural noncompliance.
No violations or deviations were identified.
7.
Monthly Surveillance Observations (61726)
The NRC d'spectors obse,ed selected portions of the performance of the TS-reg,
- surseillance testing on safety-related systems and components.
The NRL i..spectors verified the following ite.1s during the testing:
Testing was perforced by qualified personnel using appi,,ved
procedures.
Test instrumentation was calibrated.
- The TS limiting conditions for operation were met.
- Removal and restoration of the affected system and/or corrponent were
accomplished.
Test results conformed with TS and procedure requirements.
- Test results were reviewed by personnel other than the individual
directing the test.
The NRC inspectors observed the following surveillance test activities.
The procedures used for the test activities are noted in parenthesis:
Monthly (test of the A Channel of the safety injection actuation
signal ST-ESF-2)
Monthly (test of the A Channel of the containment spray actuation
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signal ST-ESF-4)
Monthly test of the A Channel of the recirculation actuation signal
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _. __
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A discussion of the surveillances observed is provided below:
On September 8,1988, the NRC inspectors observed the performance of Procedures ST-ESF-2, ST-ESF-4, and ST-ESF-13 for the A Channel of the
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engineered safeguards features system in their entirety. These l
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surveillance tests verified the proper operation of the safety injection,
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containment spray, and the recirculation actuation circuits.
The NRC inspectors observed the tests oxecuted by a licensed operator. The l
procedures were written in a manner that enabled the operator to follow
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them exactly. The operator demonstrated knowledge and familiarity with
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the systerrs which he was testing and coordinated the execution of the
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tests with the nonlicensed auxiliary building operator in a safe and i
efficient manner.
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During the performance of ST-ESF-2, the hRC inspectors independently verified that the proper lockout relays tripped follwing the introduction
'
of a low pressurizer pressure test signal into the system.
This independent verification was done by reviewing the printout of the l
emergency response facility (ERF) computer and verifying correct
'
indication of the matrix on the ERF computer display. The NRC inspectors l
also independently verified that the diesel ssquencer sent actuation signals to the appropriate equipment within the required time frames as
'
specified by the procedures.
No violations or devittions were identified.
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8.
Security Observations (71881)
l The NRC inspectors verified the physical security plan was being l
implemented by selected observation of the following items *
I The security organization was properly manned.
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Personnel within the protected area (PA) displayed their
identification Ladges.
l Vehicles were properly authorized, searched, and escorted or f
controlled within the FA.
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Persons and packages were properly cleared and checked before entry I
into the PA was pemitted.
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The effectiveness of the security program was maintained when
security equiprent failure or impatrrent required compensatory reasures to be employed.
The PA barrier was raintained and the isolation zone kept free of
transient raterial.
The vital area barriers were traintained and not cof"promised by
breaches or weaknesses.
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Illumination in the PA was adequate to observe the appropriate areas
at night.
Security monitors at the *econdary and central alam stations were
functioning properly for asse:ww:nt of possible intrusions.
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t No violations or deviatians were ident)fied.
9.
Radiological Protection Cbservations (71?091 The NRC inspectors verified that selected ac'.ivities of the licensee's
radiological protection program were implemenad in conformance with the facility policies and proced6res and in complia3ce with regulatory requirenents. The activities listed below were observed and/or reviewed:
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HP supervisory personnel c1nducted plant *,ours to check on activities
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i in progress.
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Radiation work pemics contained the appropriate infomation to
ensure work was performed in a safe and controlled manner.
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Personnel in radiation controlled areas (RCA) were wearing the
required personnel monitoring equipment and protective clothing.
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Radiation and/or contaminated areas were properly posted and
controlled based on the activity levels within the area.
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Fersonnel properly frisked prior to exiting an RCA.
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I During this inspection period, the NRC inspectors 'dentified the items listed below:
During a tour of the auxiliary building on ',eptenber 9.1988,(PING)
{
the NRC t
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inspectors noted that the particulate, iod,ne, and noble gas monitor in Room 69 was alaming.
The NRC inspectors imediately lef t l
the area and notified the enduty HP techr.ician. Approxinately
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10 ninutes later, the HP technician arrived at Room 69. The
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technician confinned that the alam was valid and infonted the NRC
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inspectors that the alam was caused by a tenperature inversion that was in pro?ress.
The NRC inspectors reviewed the chart recorder for the PING monitor
,
and noted that the monitor had been above its alam setpoint for approxinately 30 minutes. The NRC inspectors also noted that ten individuals had passed through Room 69 but had not contacted the HP technician.
The NRC inspectors expressed the concerns listed below f
to the licensee:
l l
The HP technician was slow in responding to the alam j
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notification by the NRC inspectors. Although the HP technician
was sure the alam was due to a terperature inversion, there i
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existed the potential that the alann was due to other
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conditions. Had another condition existed, the slow rc:ponse would have c used the affects of tna condition to be worse than
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they could have been.
Since a number of individuals obviously Mre aware that the PING
monitor was alarming and took no ection to notify HP of evacuate the area, personnel have become desensitized to PING monitor alams.
The licensee has historically had problems with raintaining the P!NG monitors so that they alem only when
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appropriate.
Because of the problems encountered with the alams, the monitors have falsely alanned on many occasions and
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have stayed in the alarm conditicn for days. Due to historical
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problems, individuals have become desensitized to the need to
notify HP when the monitors alam.
If personnel become desensitized with respect to the PING monitor alams, then it is I
probable that they will also becoce desensitized about other
HP-related alams such as the area radiation monitors.
i The concerns related to operation of the PING monitors was discussed
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t with licensee management, t.icensee management concurred that actions l
needed to be taken to ensure that appropriate response is provided
,
when the PING monitors alam.
By the end of this inspection period.
>
the Itcensee had not provided a specific action plan to address the
'
concerns. This item remains open pending a response to the concerns
,
discussed above with licensee management and a review of the proposed actions by the NRC inspectors.
(285/8832 04)
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b.
During the course of the month, the NRC inspectors noted that the licensee had changed its method of posting local radiological hot spots in the auxiliary building. A hot spot is a localized area L
where radioactivity levels exist which are significantly higher than
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the general area. The hot spots require posting in accordance with i
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Procedure VII-9-25. "Radtation Hot Spot Verification.' of the
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Radiological Protection Manual.
Tha NRC inspectors noted that the fcnner self-adhesive stickers had been replaced by hanging plastic tags for identification of hot l
spots. The stir.kers were, by their nature, pasted to the hot spot.
l However, the new tags are yellow and magenta, with the international i
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synbol of radioactivity the trifoil, on one side and undesignated i
white plastic on the other. Therefort, it was possible for an
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individual to approach a hot spot on the reverse side of the tag and I
not be aware of the increased radiation level.
On September 22, 1988, the hRC inspectors brought this observation to the attention of the plant health physicist. The plant hP agreed I
with the NRC inspectors that the postir0 technique could be improved.
l On a tour of the auxiliary building on Septerber 23, 1988, the NRC i
inspectors noted that questionable tags had been rehung in a manner i
rNktng the hot spot visible from a 360-degree periphery.
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l In followup on this iten, the NRC inspectors examined
Procedurs VII-9-25 "Radiation Hot Spot Verification." The l
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examination revealed that administrative controls were in place for
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posting and surveillance of radiological hot spots. However, there i
were no specific instructions pertaining to the proper method of
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posting hot spott. The NRC inspectors addressed this concern to the
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plant HP who agreed that in view of the problems identified with the new tags, specific instructions were warranted.
The plant HP agreed to have Procedure V11-9-25 revised to include instructions on how to post hot spots. The issuance of this procedure revision is an open
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item. (285/8832-05)
i No violations or deviations were identified.
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10.
In-office Review of periodi and Special Retorts (90713)
i In office review of periodic and special reports was performed by the NRC resident inspectors and/or the NRC project engineer to verify the
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following, as appropriate:
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Correspondence included the information required by appropriate nRC
l requi rerents.
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Test results and supporting infomation were consistent with de:Ign
predictions and specifications.
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Detemination that planned corrective actions were adequate for
resolution of identified problems, j
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Detemination as to whether any infomation contained in the
j corrcspondence report should be classified as an abnorral occurrence.
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Correspondence did not contain incorrect, inadequate, or incomplete
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I infornution.
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The NRC inspettors reviewed the following correspondence:
August Monthly Operating Record, dated Septerter 15, 1988 f
Special Report on !rcperability of Fire Barrier, dated September 15
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1988
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Special Report on Heated Junction Thermocouple Inoperability, dated
Septecber 14,1GS3
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Special Report on Fire Barrier Seals 26-N-24 and 69-F-136, deted
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Septerter 23, 1988 j
i Special Report en Fire Dour Locksets and Coor Modifications, dated
September 20, 1988 i
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Clarification of Information Relating to tne Op6*ation of the
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Internals Vibration Monitoring System, dated September 19, 1988 f
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Special Report on Inoperability of Fire Barrier, dated September 2,
1988
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TS Amendment Request for Cycle 12 Operation, undated j
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j Response to *1RC Bulletin 88 05, dated September 9, 1988
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TS Amendment Renuest for Section 2.7, "Electrical Systems," dated l
September 9, 19&\\
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I No violations or deviations were identified.
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11. Review of Generic letter 88-03 (92703)
I i
On February 17, 1988, the NRC issued Generic Letter (GL) 88 03, i
"Resolution of Generic Safety Issue 93 Steam Binding of Auxiliary) system Feedwater Pumps." This GL noted that the auxiliary feedwater (AFW l
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was vulnerable to backleakaga from the steam generator into the AFV system
i via system check valves. The GL requested that the licensee implement
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i procedures for monitoring the AFW piping temperatures and for restoring I
l the AFW pumps to an operable status if steam binding occurs.
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In a letter dated May 25,1988, the licensee provided a response to the
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issues identified in GL i;8-03. A discussion of each item in the licensee's response is provided below:
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Surface temperature indtcators have been installed on the discharge l
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- iping of both AFW pumps.
i The temperatu6 e indicators are monitored once each 8-hour shif t and
recorded on an operations log.
Procedure 01-AFW-2, "Auxiliary Feedwater System Operation (Special
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Procedures)." was issued to provide instructions to operations i
personnel on the actions to be taken in the event the AFW pumps became steam bound.
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j The NRC inspectors reviewed the actions taken by the licensee, as
described above. Based on this review, it appeared that the licensee had i
i adequately addressed the issues identified in GL 88-03.
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f No violations or deviations were noted.
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12. Review of NRC Bulletin 88-05 (92703)
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j In paragraph 12 of hRC Inspection Report 50-285/88-23, the NRC inspectors j
provided details of a review perfomed by the licensee as required by NRC r
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Bulletin 88-05, "Nonconfoming Material Supplied by Piping Supplies, Inc, j
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j at Folsem New Jersey, and West Jersey Manufacturing Company at
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Williamstown, New Jersey." During the review perforned by the licensee, i
ten flanges were identified that had been supplied by the West Jersey
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Manufacturing (WJM) Company.
Six flanges were installed in the waste disposal system, and four blind flanges were installed on containment
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.
On September 9,1988, the licensee submitted a response to NRC
l Bulletin 68-05 to provide details of their review of flange purchases nade j
at the FCS. Based on the review of all purchase orders, the licensee did l
j not identify any flanges other than the ones discussed in NRC Inspection
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Report 50-285/80-23.
t The licentee's response also provided information on testing that had been I
done to verify acceptability of the flanges. The licensee tested eight of
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ten flanges. Two flanges are located inside containment and were not
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accessible for testing during power operation.
The resu'ts of the testing l
indicated that the flanges were acceptable for continued system operation, g
Acceptability was based on the preparation of a calculation which was
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developed using the lowest tensile strength value reported to date in the
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j industry and the maximum system pressure exerted on the flanges.
This
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t value used in the calculation was 42 ksi. All flanges tested were found l
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to have a tensile streng h of 65 to 68 ksi.
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j The NRC inspectors reviewed the submittal made by the licensee.
It appeared that the licensee had adequately addressed the concerns identified in NRC Bulleein 88-05.
This bulletin remains open p( iding a detailed review of the licensee's submittal by NRR.
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No violations or deviations were identified.
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13.
Infonnation on Containment Temperatures (71707)
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j Ouring this inspection period, the hRC inspectors reviewed cata on l
containment temperatures over a 52-week period beginning in August 198i.
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The data were provided by the licensee.
The data review was done per NRC
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Temporary Instruction 2515/98. The purpose of this inspection was to obtain a containment terperature profile and to detemine its effect on i
the enviromental qualification of equipment, in particular, electrical
,
j insulation.
It should be noted that the licensee already had its own l
program in place to obtain a temperature profile of the containment.
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l Based on the review of the data provided by the licensee, the NRC
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inspectors found that the arithnetic rean temperature at the 1045-foot i
j elevation for the 52-week period. August 1987 through July 1988 is 86.9'F.
l This average is above the yearly average terperature of 85'f which the
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l licensee uses in its equipment environmental qualification program when I
r calculating the remaining qualified lifetire for equipa mt inside the i
J containment.
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The 85'F temperature was obtained by the licensee in 1982 from individual l
temperature measurements taken at 16 containment locations and averaged arithmetica11y.
The readings were taken on a single day at 100 percent power.
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The licensee contracted the engineering firtn of Sargent and Lundy to evaluate the basis of the temperatures currently used to qualify EEQ
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devices under normal operating conditions. The Sargent and Lundy study is
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not yet complete but the data suggests that everage annual ambient t
temperatures could be in excess of 85'F.
Sargent and Lundy states that
the data for tre 52-week period August 1987 through July 1988 is not of a l
nature to either conclusively support or deny the assertion that average l
l annual temperatures greater than 85'F are occurring.
The data obtained in t
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the 52-week study is uncontrol;ed. That is, no written procedures were used to record it and no regulated calibration was documented.
Therefore, the Sargent and Lundy study recortrrended that the licensee j
j establish an ongoing temperature monitoring program to obtain an adequate f
t dtta base. The Itcensee has initiated this program.
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t The licensee demonstrated, however, in Operations Support Analysis
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Report 85-67, that a qualification temperature of 90*F could be used with
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no required rebuilding or maintenance of qualified equipment.
Based oa l
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J the review of this document, it appears that an adequate targin of safety
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j ca.ists in using a qualification temperature of 85'F.
Therefore, an j
j average annual ambient temperature of 86.9'F should not significantly i
degrade qualified lifetire of equipment.
This is the same conclusion (
)
reached earlier in NRC Inspection Report 50-285/85-26, Details.
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Section f.0, during followup on environmental qualification issues.
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All the preceding information was obtained by the NRC inspectors for i
6nalysis by NRR.
Subsequent to an assessment of the locations in which i
the temperature sensors are placed in containrent, the data will be
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forwarded to hRR for review. This will be considered an open item pending hRR's review.
(285/8832-06)
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No violations or deviations were identified, c
l 14.
preparation for Refueline (60705)
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The NRC inspectors reviewed selected procedures to verify that the f
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technical and ad91nistrative requirements provided in the procedures l
i adequately inccrporated controls and requirements stated in the TS. codes, I
and standards,. The procedures revicwed were evaluated to verify that controls had been established for the following eierents:
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l Fuel bandling, transfer, and core vertftcation
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f j
Inspection of irradiated fuel
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Preoperational checkout of refueling equipment
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_ _ _ _ _ _ _ _ _ _ _.
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Handling and troverrent of the reactor core internals
Establishment of administrative measures for activities such as lines
of supervision, shif t tranning, radiation rionitoring, comunications,
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and training Invoking of casualty procedures
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Pequirements for work stoppage, delays, and holds
The NRC inspectors reviewed the following procedures for the elements listeo above.
Number Title Revisica SP-CEA-3 Procedure for rEA Unloading. Inspection, and
Storage of Individual CEAs OP-11 Reactor Core Refueling Procedure
SP-FE-1 Visual Inspection of Irradiated fuel
MP-FH-2 Refueling Equiprent Preoperational
Maintenance MP-HE-1 Polar Crane Annually or at Pefueling
Inspection MP-RC-7-2-A Upper Guide Structure and ICI Plate Removal
The NRC inspectors witnessed activities related to the receipt, inspection, ard storage of new fuel and CEAs.
The results of this inspection are docurente.d in NRC Inspection Report 50-285/E8-27.
The procedural review perforced by the hRC inspectors found the rafueling-related procedures to fully reet the requirerents specified in ANSI 18.7-1972 and Regulatory Guide 1.33, Appendix A.
This is the eleventh refueling outage for the licensee and it is apparent that the procedures involved in the refuelir.g proce's hne received adequate attention in updating and revisions, ho violations or deviations were identif en.
If;.
Exit Interview The NRC inspectors ret with Mr. K. J. Morris (Division Manager-Nuclear Operatiens) and other rerbers of the licensee staff at the end of this inspection. At this neeting, the NRC inspectors sumarized the scope of the insfection and the findings.