IR 05000293/1988031
| ML20205P723 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 10/24/1988 |
| From: | Blough A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20205P715 | List: |
| References | |
| 50-293-88-31, NUDOCS 8811080272 | |
| Download: ML20205P723 (27) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No.:
50-293 Report No.:
50-293/88-31 Licensee:
Boston Edison Company 800 Boylston Street Boston, Massachusetts 02199 Facility:
Pilgrim Nuclear Power Station Location:
Plymouth, Massachusetts Dates:
August 29, 1988 - September 26, 1988 Inspectors:
C. Warren Senior Resident Inspector C. Carpenter, Resident Inspector T. Kim, Resident Inspector J. Lyash, Project Engineer R. Barkley, Reactor Engine <r G. Napuda, Lead Reactor Inspector i
E. Trottier, NRR
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T. Dragoun, Senior Radiation Protection Inspector Approved by:
hk 4 D~d(P C Randy Blgssh, Chief Data Reactor Projects Section No. 3B Olvision of Reactor Projects Areas Inspected:
Routine resident inspection of plant operations, radiation protection, physical security, plant events, maineenance, surveillance and out-age activities.
Principal licensee management representatives contacted are listed in Attachment I to this report.
Results:
Violation: Not all personnel radiation exposure history reports were issued as required daring 1986 and 1987.
Unresolved Item:
The inspector walkdown of the control room bigh efficiency air filtration system identified seural apparent discrepancies in both opera-ting and turveillance procedures.
This item is unresolved pending review of the licensee's response (UNR 88-31-01, Section 3a).
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TABLE OF CONTENTS
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i 1.
Summary of Facility Activities..............................
2.
Followup on Previous Inspection Findings Violations, Unresolved Items, Inspector Follow Items (Module Nos. 92701 and 92702).............................
3.
Noutine Periodic Inspections (Module Nos. 71707, 71709, 71710, 61726, 62703, 92702, 02703, 71881 and 37700).......
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System Alignment and Surveillance Testing Inspection...
b.
Plant Maintenance and Outage Activities................
c.
Licensee Actions in Response to NRC Bulletin 88-05.....
d.
Licensee Actions in R(sponse to NRC Bulletin 88-07.....
4, Receipt of Licensee Keys (Module No. 92701)................
5.
Allegation Regarding Quality Control Inspector Quslifications (Module No.
92701).........................
6.
Managenent Meetings (Module No. 30703)......................
Attachment I Persons Contacted
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f DETAILS
1.0 Summary of Facility Activities The plant was shutdown on April 12, 1986 for unscheduled maintenance. On July PS,1986, Boston Edison announced that the outage would be extended to include refueling and completion of certain modifications.
The ieactor core was completely defueled on February 13, 1987 to facili-
tate extensive maintenance and modification of plant equipment.
The licensee completed fuel reload on October 14, 1987. Reinstallation of the
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reactor vessel internal components and the vessel head was followed by i
complation of the reactor vessel hydrostatic test.
The primary contain-
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ment integrated leak rate test was also completed during the week of l
December 21, 1937 During this period, the licensee performed routine
maintenance and surveillance tests.
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NRC inspection activities during the report period included: 1) a review
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of the licensee's corrective actions on previous NRC inspection findings during the week of September 5, 1988, 2) a review of the licensee's cor-rective actions on previous electrical inspection findings during +.he week
of September 12,1988 and 3) review of the licensee's corrective actions on Integrated Assessment Team Inspection (IATI) report findings.
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The NRC issued the IATI Inspection Report 50-293/88-21 on
September 7, 1988.
The IATI was conducted on August 8-24, 1988. A full i
committee meeting of the Nuclear Regulatory Commission's Advisory Commit-tee on Reactor Safeguards (ACRS) was held on September 8, 1988, in Bethesda, Maryland to discuss the proposed restart of the Pilgrim Nuclear i
Power Plent.
The ACRS recommendation to the Commission on the proposed restart of the Pilgrie plant was issued on September 14, 1988.
On Septembe.- 11, 1988 Cynthia Carpenter assumed the posttion of NRC Resident Inspector at Pilgrim, replacing Jeffrey Lyash who was proLoted to
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Proje:t Engineer, Reactor Projects Section 3C, Region 1.
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2.0 Followup on Previous Inspection Findings Violations (Closed) Violation (87-29 01). Reactor Operater was Found Inside a High Radiation Area Without the Required Radiation Survey Meter.
The licensee i
investigation determined tha. the event was an isolated case of personal l
error. Co rective action described in the licensee response letter dated September 23, 1987 (BEco Ltr. #87-154) is complete and satisfactory.
(Closed) Violation (87-33-01). Failu*e to Adequately Review Plant Design Change _(PDC) 86-70 and Temporary Procidure (i~P)87-128 for the Standby 34s Treatment System. Lictnsee corrective actions were previously reviewed in inspection reports 50-293/87-41 and 50-293/88-03.
The ifcensee formed an indeoendent three-man tean: composed of senior consultant engineers to
review modification mar,agement test procedures for plant design changes (PDC) implemented for the duration of refueling outage No. 7.
The intent of the review was to assure that the testing met the requirements of the
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plant design change. Guidelines were established outlining how to conduct
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reviews of the plant test procedures.
In addition, upon completion of
these reotews, the itcensee evaluated the results of the three-man team's independent review ef fort and determined that, depending on the number of
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modifications being implemented in future outages, that the nuclear organitstion should consider setting up a special team to review pre-operational tests for technical adequacy for the outage. Licentre actions
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regarding t;its violation have been found to be complete and satisfactory.
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LClosedl Violatio in87-40-01). Failure to Post a Radiolgical Violatirn Per 10'CFR7;9.11. The lic'ensee subsequently posted the violation as required.
To prevent recurrence; the responsibility for posting is now assigned to the Regulatory Compliance Response Coordinatcr by Work Instruction #371.
In addition, the responsibility for providing an adequate number of copies of the violation documents f s assigned to the Document Control Supervisor by Record Management Group Work Instruction #2.31.
Licensee corrective action is complete and satisfactory.
(Closed) Violation (87-50-04). Failure to Adhere to an Established Radia-tion Work Fermit. A radwaste worker entered a High Radiation Area without the required RWP, anti-contamination clothing or HP technician coverage.
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Licensee corrective Action described in a letter dated March 24, 1938
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(BECo Ltr. #88-058) is complete and satisfactory.
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(Closed) Violation 87-50-07, Failure to Properly Preplan and Perform Maintenance in Accordance with Sufficiently Detailed Instructions.
During inspection 50-293787-50 the inspector found that the licensee's mainten-ance process did not contain provisions for the generation and use of suf-ficiently detailed job specific work instructions.
As a result, replace-ment of an electrical relay coil was not adequately preplanned and docu-mented, and performance of the activity caused an engineered safety fea-ture (ESF) actuation.
Licensee response to the violation was documented by letter dated March 24, 1988.
The short-term corrective actions dis-cussed in this letter were adequat9.
The licensee subsequently imple-mented a major change to the maintenance process This change included establishment of a "Maintenance Work Plan" system which provides signifi-cantly improved control.
This new process was reviewed in detail during Integrated Assessment Team Inspection (IATI) 50-2S3/88-21 and was found to be effective.
Based on the licensee's response to the violation, and the inspection performed during the IATI this itein is losed.
(Closed) Violation (87-57-01), Failure to Adequately Control Access to Locked High Radiation Areas. On three occasions, ths doors to Locked High Radiation Areas were found to be unlocked and unatterMed. During inspec-tion 50-293/88-22 it was determined that the licensee had implemented the corrective actions described in the April 15, 1988 response letter (BECo Ltr. #88-070) except for the additional RWP controls specified. In April 1988 the licensee developed a list called "Standard Requirements for Entry
- Locked High Rad Areas" which is attached to all topropriate RWPs. The front of the RWP is stamped to reference the attacned requirements.
In addition, procedure 6.1-012 "Access Control to High Radiation Areas" was revised to incorporate the standard entry requiremer.ts.
Licensee action on this matter is complete and satisfactory.
(Closed) Violation (88-07-01), Failure to Establish Adequate Instructions and to Perform Adequate Technical Reviews of Plant Design Changes. During inspection 50-293/88-07 the inspector ider.tified two instances of inade-quate instructions contained in approved plant design change (PDC) pack-ages.
Licensee response to this violation is documented in their letter dated June 10, 1988.
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In the first example instructions for installation and testing of a new computer point contained in PDC 83-51D resulted in unnecessary engineered safety feature (ESF) actuations. The effect of breaking a common neutral circuit during the activity was not recognized. The licensee immediately stopped all ongoing computer point tie-in work and revised instructions in the PDC where required.
An Engineering Service Request (ESR) was issued to evaluate the feasibility of impismenting permanent modifications to remove the "daisy-chained" neutral circuit.
In the interim, caution y
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signs have been installed in the subject panels. In addition, the licen-
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see implemented a major revision to the maintenance request (MR) process in June, 1988.
This revision requires the use of more detailed, job specific work instructions.
Guidance on development and review of these instructions was also created.
The effectiveness of the new MR process was evaluated during Integrated Assessment Team Inspection 50-293/88-21 and found to be acceptable.
The licensee's initial followup to the ESF actuations caused during the computer point tie-in was not effective.
The critiques held to discuss
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the incident did not involve all appropr'. ate personnel and did not iden-tify the cause of the actuations until pointed out by the inspector. The licensee subsequently refined the incident critique process and formalized it through issuance of procedure 1.3.63, "Conduct of Critiques and Inci-dent Investigations".
The inspectors have attended a sample of recent critiques and noted a r..a r ked improvement.
The effectiveness of the t
critique process should continue to improve as experience with its imple-mentation is accumulated.
The second exampic cited in the violation was the failure to establish an
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adequete procedure for post-modification testing of the blackout diesel
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generator.
Daring review of procedure TP 88-09, "Electrical Plant Line-Up i
for Blackout Diesel Generator Load Test", the inspector identified that l
performance of the test as written would have resulted in an unanticipated stari af the "B" emergency diesel generator (EDG).
The licensee immedi-ately suspen ed performance of the test, reviewed the procedure, and iden-r i
tified and currected a number of deficiencies.
The licensee subsequently revised procedure 1.3.4, "Procedures", to require inclusion of a descrip-
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tion section in all test procedures which will assist in determination of plant response.
The Modification Mana ment Division revised the "Manual of Work Processes" to require specif c review of test procedures for potential ESF actuations.
In addition the Systems Division has develuped I
a checklist for use during procedure reviews whien includes consideration of potential ESF actuations.
Subsequent to licensee submittal of their violation response, a dedicated procedures development group was estab-lished and a procedure validation prucess was implemented. These enhance-ments will provide additional checks.
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The licensee has exper'enced a high number of ESF actuations during the past year.
In response, a multidisciplined task force was established to evaluate root causes and to recommend corrective actions.
The licensee was requested to submit a report describing the results of this effort for i
NRC review as noted in inspection report 50-293/88-25.
Followup of the licensee's response in this ares will be continued under existing unre-solved item 88-25-02.
Based on the inspector's review and an existing item 88-25-02, this item is closed.
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(Clesed) Violation 88-12-02, Inadequate Design Control and Preoperational testing for Reactor Water Level Gauge Installations LI-263-59A & B.
The licensee's response to the violation, dated July 8,1988, stated that the root cause of the improper installation of the gauges was an incorrect installation drawing supplied by the vendor.
The drawing was in error because the vendor reversed the high and low pressure taps on the instru-ment to meet the licensee's design specifications and did not identify that change on the drawing. In addition, the instruments were erroneously omitted by the licensee from Temporary Procedure TP 86-188, "Recalibration cnd Test of Proportional Amplifiers PA 640-3A & B and Various Reactor t
Water Level Transmitters." Inclusion of the instruments in the TP 86-188 would have identified the deficiency in the installation of these
instruments.
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To correct the misinstallation, the licensee issued maintenance requests 88-45-47 & -48 to reverse the connections on the instruments.
In addi-tion, Field Revision Notice 85-07-126 was issued to correct the drawing error.
Procedure 2.1.15 was also revised to include a channel check of tho instruments on a daily basis.
The inspector also reviewed the licen-seee's investigation and corrective actions as documented in Potential Condition Adverse to Quality (PCAQ) report 88-013.
The licensee's inves-tigation indicated that this misinstallation was an isolated event. This conclusion is supported by a Nuclear Engineeritig Department review of instrument modifications and initial installatter tests performed during
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the outage. That review did not identify any other instrument misinstal-l lations in the plant.
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The inspector did note that in the violation respo9se, the licensee indi-cated that there are level indicators available on the Anticipated Trans-1ent Without Scram ( ATWS) panels 2277 and 2278 that also indicate reactor water level, mitigating the safety significarce of an event where LI-263-59A and B are unavailable.
While those level indicators do provide an indication of reactor water level, the inspector stated that they do not appear to significantly mitigate the need for LI 59A and B since they m.y not be operable in the event of a 10 CFR 50 Appendix R type fire (which was the original reason for the installation of LI-263-59A and B.) This
'wa s discussed with the licensee.
The inspector had no additional
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q ue s t 's on s.
Unresolved Items (Update) Unresolved Item (86-21-03), Inadequate HFCI Logic System Func-tional Tests, this item was last updated in inspection report 50-293/
85-19.
Previous NRC reviews conducted in this area verified the ef fec-
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tiveness of licensee actions to establish technically adequate logic sys-tem functional tests (LSFT) for all plant systems. Additionally, a sample of LSFTs was reviewed during NRC Integrated Assessment Team Inspection (IATI) 50-293/88-21. As a result of these reviews all technical concerns wert resolved and only five administrative items remained.
During the current period, the inspector evaluated the status of these administrative items.
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Evaluation of procedural controls implemented to assure review of l
plant modifications so that compliance with surveillance requirements i
for LSFT is mafr.tained.
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Nuclear Engineering Department and Nuclear Operations Department pro-cedures describing the process for control of plant modifications clearly address the identification and revision of affected proced-
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Modification Management instructions provide for the tracking l
and closecut of modifications related procedure changes.
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2.
. Evaluation of procedural controls to assure a LSFT review is per-formed for procedure changes impacting LSFT surveillances.
Procedure 1.3.4, "Procedures", and 1.8, "Master Surveillance Tracking Program" establish the review and approval process which is applied to each Procedure Change Notice (PCN).
Responsibility for the var-ious technical reviews is clearly established. Review of each PCN by the appropriate systems engineer specifically for impact on LSFT is required by procedure 1.8.
A computer data base has been developed to assist in this effort. Operation of the data base is controlled by instruction SI-TC.3.2.13, "Control of LSFT/SAA Data Base". Updit-
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ing of the data base for each PCN is required by procedure 1.8.
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inspector concluded that a strong program for maintenance of LSFT technical adequacy in the event of procedure change notices exists.
3.
Review the licensee plans to implement long-term recommendations identified in the final report.
The contractor LSFT final report contained a number of long-term recommendations.
These recommendations generally address drawing inaccuracles and possible procedure changes and enhancements.
The recommendations were considered desirable but not required.
On September 6, 1988, the licensee's Nuclear Engineering Department forwarded by memo a summary of these items to the Plant Manager for followup.
The inspector reviewed this memo and noted that while many l
of the items are minor in nature several appear of enough signifi-
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cance to warrant additional review.
The licensee informed the inspector that all drawing discrepancies had been dispositioned and
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that many of the other recommendations had been implemented. How-
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ever, no individual was assigned to ensure completion of the effort and no status of individual items was available. The licensee stated
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that a review would be initiated and a status of each item provided.
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Licensee disposition of Final Safety Analysis Report (FSAR) commit-ments not presently tested, as identified in Table 4 of the final contractor report.
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The contractor report identified five cases in which system functions
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described in the FSAR were not included in LSFT.
It was the contrac-tor's position that testing of thr'e features was not required by Technical Specifications (TS) but snould be evaluated.
Engineering evaluations of these five items were conducted by the licensee and appropriate surveillance testing was established.
5.
Verification of surveillance testing frequency for LSFT on the Master Surveillance Tracking Program (MSTP).
The licensee performed a review of surveillance test frequencies a
specified in the MSTP and concluded that they are presently in I
accordance with the TS.
The MSTP data base and the controls applied to it were reviewed during IATI 50-293/88-21 and found to be ade-quate. One outstanding question identified during the IATI concerned the justification for performance of LSFT of the Reactor Core Isola-i tion Cooling System once per operating e.yele vice semiannually. This was identified for followup under urr.esolved item 88-21-01.
During j
the current period the inspector tvaluated a sample of LSFT surveil-i l
lance frequencies as contained *.i the MSTP and found them to be con-i sistent with TS requireme.its.
The inspector had no further l
questions.
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This item will remain open pending licensee response to the inspectors
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questions concerning the disposition of the long-term recommendations as i
discussed in item 3 above.
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l LClosed)UnresolvedItem(86-24-01), Licensee to Provide Definitive Di,ec-
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tives to Assure that Vendor Technical Information is properly Forwarded for Centralized Incorporation into the Master Information Files.
DWing inspection 50-293/86-24 the inspector noted that development of the vendor technical information (VTI) review and control remained incomplete.
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current VTI program is no' centralized, in that the program is implemented
by several groups. The O. rating Experience Review Program which provides
control of information generated by INPO and General Electric is main-tained by the Systems Division.
The Regulatory Affairs and Programs i
Division controls information generated by the NRC.
Vendor information from other sources such as individual suppliers is reviewed by various
portions of the line organization. While the program is not centralized,
it appears that an adequate program is currently in place to ensure VTI is
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properly reviewed and incorporated.
The licensee committed in their i
Restart Plan (Appendix 10, Issue 03-940-01) to consider key elements of
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I the Nuclear Utility Task Action Committee Vendor Equipment Technical
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Information Program (VETIP), including a centralized approach to imple-menting the VETIP. Their schedule is restart plus 240 days. The inspec-t tor had no further questions.
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(Closed) Unresolved Item (86-43-01), Licensee to Develop Procedures for Use and Training of Contractors.
Procedures PNPS 1.5.3,
"Maintenance
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Requests", and PNPS 1.5.3.1, "Maintenance Work Plan", are fully applied
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to contractor performed activities. Additionally, contractors performing maintenance on plant equipment are receiving training in accordance with the M1-Maintenance Training Program that is given to all station mech-anics.
Further, the Construction Maintenance Division has a series of instructions that include contractor control provisions such as Instruc-tion Number 11. "Contracts Administration"; Number 14, "Time and Material Contracts"; Number 20, "Daily Inspections"; and Number 29, "Standards and Practices".
The inspector had no further questions.
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(Closed) Unresolved Item (87-27-07), Potential Violation of Employee
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Radiation Exposure Reporting Requirements.
Inspection report 50-2937 F27, documents an allegation that an individual had not received a
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report of his radiation exposure during employment at Pilgrim upon his
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termination in 1986.
Additional requests for radiation ecoosure records by the alleger were unanswered.
The inspection report further documents that the licensee's Radiation Protection group evaluated the allegation and determined that in excess of one hundred similar failures to provide the required reports had occurred.
The licensee indicated at the exit meeting that the backlog of exposure histories would be reviewed and the required reports issued.
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Subsequent licensee review identified approximately 400 owrdue termina-tion exposure history reports, beginning in early 1986. Between' April and August 1988 the licensee issued all those deliquent reports. The licensee also determined in August 1988 that approximately 15 former BECo employees terminating employment between October 1987 and April 1988 had also not received the required reports.
These reports are currently being issued.
10 CFR Parts 19 and 20 require that the licensee supply exposure records upcn an individual's temination or upon request.
Furthermore, procedure SI-RP.1405, "Personnel Terminations", requires that temination letters be issued within 30 days af ter the final exposure of the individual has been dete rmi ned, or 90 days af ter the termination date, whichever is earlier.
Therefore, failure to provide the required reports in a timely manner constitutes a violation of 10 CFR 19 and 20 and procedure SI-RP.1305.
The inspector sampled a portion of employee termination files and deter-mined that the licensee is now providing exposure history reports as required.
Additionally, requests for exposure hist;ry also appear to be provided as required.
The licensee now utilizes a computer tracking system to log all requests for exposure history and dates of employee termination f rom Pilgrim.
Corrective actions by the licensee appear to have been adequate to address this issue.
The licensee has identified those individuals who did not receive exposure history reports and has issued the required reports. Curren. reports are being issued in a timely manner.
The inspector has determined that the licensee's corrective actions are acceptable. Therefore, no written response to this violation is necessar :
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(Closed) Unresolved Item (87-27-01), FSAR and Annual Report of Changes Made to the Facility Under 10 CFR 50.59 Did Not Include the Change Made to the Core Spray Injection Valve M0-1400-25 Stroke Time.
The inspector reviewed FSAR section 7,4.3.4.4, Revision 9, and noted that the valve stroke time for MO-1400-25 reflected the new value.
In addition, the 1987 annual report of changes, tests and experiments performed at Pilgrim included the change to the valve stroke time for MO-1400-25, along with several other changes made to the facility prior to 1987, which had not been reported.
These changes were identified due to refinements made to the 50.59 reporting process as a result of inspection 87-27, None of the changes noted in the report represented unreviewed safety questions nor were of safety significance.
(Closed) Unresolved item (87-47-01), Licensee to Oetermine the Root Cause of the Failure of Torus Vent Line Main Exhaust Valves A0-5042A and A0-50428 to Pass Local Leak Rate Testing (LLRT).
The licensee determined that the cause of the failure of A0-5042A and A0-50428 to pass (LLRT) was due to the introduction of condensate demineralizer resin into the torus vent line.
The resin accumulated on the seats of the valves, preventing the valves from properly sealing upon closure.
The resin apparently had been introduced into the vent line years earlier from the condensate demineralizer vent to the Reactor building Exhaust Plenum.
The resin traveled by gravity into the torus vent line, which is connected to the Reactor Building Exhaust Plenum. Over time, some resin then became lodged
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in the valves, which are located on a vertical section of the torus vent line.
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The licensee has taken action to permanently correct this problem by cut-ting and capping the condensate demineralizer vent to the Reactor Building Exhaust Plenum. The licensee also cleaned the seats of valves A0-5042A and 5042B and reperformed the LLRT with acceptable results. The inspector had no further questions, i
(. Closed) Unresolved Item (87-50-06), Review Improperly Assembled RWCU I_n s t rumen t Flow Snubbers and piping Design.
During followup of seve'al
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spurious Reactor Water Cleanup (RWCU) high flow isolations, the licentee identified that flow snubbing devices installed in the instrument piping immediately upstream of the flow sensing instrument isolation valves were improperly configured.
Further investigation revealed that 56 flow l
snubbers had been installed in six safety-related systems in 1972, using l
a maintenance request (MR).
Installation under an MR without issuance of
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a plant design change (PDC) bypassed the design control process, so that l
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no clear desig' basis was established.
Inspection of the snubbers iden-I tified that the settings were not consistent, and several appeared I
l partially blocked by corrosion products. Because the snubbers are located
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upstream of the instrument isolation valves thcy were not tested during
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the instrument calibration.
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The licensee implemented an inspection program to document the as-found configuration and location of each flow snubber.
The snubbers were originally installed in-line with each Barton differential pressure switch to reduce syster noise and oscillations; thereby reducing wear of the instrument mechanical '.inkage and resultant setpoint drift. An engineering evaluation was conducted to establish the general design basis for the snubbers, and the specific configuration required for each application.
The snubbers were flushed to remove any trapped debris.
Plant Design Change 88-25 was issued to reconfigure each snubber to provide appropriate damping action based on system response time requirements and instrument characteristics.
In one application, on the main steam line (MSL) high flow sensors, the snubbers were altered to provide minimum damping and were scheduled for removal during the next outage. Since the original MSL-igh flow Barton mechanical sensors have been replaced by electronic dif-ferential pressure instruments no additional damping is needed.
Removal of these snubbers has been added to the licensee's Long Term Plan.
Because the damping action of the snubbers was reduced for instruments in the residual heat removal (RHR) system, augmented monitoring for excessive oscillation has been scheduled during the power ascension program.
In addition calibration of these instruments has beet scheduled on a quar-terly basis to detect any increased wear or o ift. The licensee will also establish measures to assure flushing of the snubbers during each refuel-ing outage.
Implementation of thest actions is tracked by PDC 88-25 and has been assigned to the Maintenance Section. The POC in conjunction with the measures described above appear
'.o resolve this concern.
The inspec-tor had no further questions.
(Closed) Unresolved Item (87-50-08). Latch Taped on Locked High Radiation Area Door.
Based on NRC review of the licensee's response to Notice oT Violation 87-57-01, this item is closed. Documentation of this review can be found in NRC Inspection Reports 50-293/88-21 section 3.5.2.6 and 50-293/88-22 section 3.0.
(Closed) Unresolved Item (87-58-01), Deficient Administrative Control Ouring the ILRT. During tha containment integrated leak rate test con-ducted on December 20-23, 1987, the licensee identified a water leak in the high pressure coolant injection (HPCI) turbine room.
It was deter-mined that increasing pressure in the torus air space had caused the sup-
.l pression pool water to back up through the HPCI turbine exhaust line and I
through the drain piping, overflowing the HpCI gland seal condenser onto the HPCI room floor.
The turbine exhaust line discharges to the torus
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through a check valve and a locked open stop-check valve. To prevent any condensation from collecting in the turbine exhaust line downstream of the check valve, piping drains any condensation to the HPCI gland seal conden-ser through a drain pot. The solenoid operated drain valves on the drain
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pot close automatically on a HPCI (Group IV) isolation signal. This is to
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provide the isolation from the torus to the gland seal condenser.
The
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licensee't investt]atier, determined that leads had been lifted in the HPCI
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isolation logic circuit in support of HPCI steam testing utilizing tem-
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porary oil-fired auxiliary boilers.
With the HPCI isolation signal by-past i, th< drain valves remained open as the drain pot was filled with the sepp *ession pool water.
The licensee subsequently relanded the leads a r' 9e drain valves closed.
,
Af**r hwing the ILRT procedure, HPCI test procedure and interviewing li
, per!)nnel, the inspector concluded that licensee review of the act, main.tenance requests prior to the ILRT was not thorough in that the lifted lew, controlled by MR 87-663 were not identified.
The MR tags were attached to the HPG isolation logic circuit inside a logic panel and thus the tags were not identified during a system walkdown nrior to the ILRT. The drain valve positions were verified by the light indications on control room panel 903 as prescribec in the ILRT procedure and not locally.
The inspector also determined that the maintenance request in this case was not an adequate method of identifyi., and tracking jumpers and lifted leads, especially for a long-term application and for com-f ponents which colld affect other ongoing maintenance or surveillance.
The licensee now has a station procedure which requires additional con-trols for jumpers and lifted leads which are covered by active mair.tenance requests. A lif ted leao's &nd jumper log is being kept in the control room to aid the operators. The inspector had no further questions.
(Cicsed) Unresolved Item (88-07-02), Evaluate Post-Work Testi, of Work Perfctmed Under the E-203 Project.
Inspection Report 50-293/88-07, sec-tion 4 G, describes in detail the licensee's E-25' program which was l
undertaken to replace wire lugs, fuse blocks and terminal strips, A
history of problems in maintenance scheduling, planning and post-work testing brought into question the adequacy of the test'ng of work per-formed ender the E-203 program.
The licensee conducted a thorough review of all maintenance requests (MR)
which performed E-203 work and developed a tes'.ing matrix which ensu. ed that testing was performed ir. all cases where wire splices were made or 6 cre than one lead was lif ted simultaneously.
In addition, a representa-tive sample was tested where work involved the lif ting of only one lead.
15e testing consisted of functional tests in the form of routine surveil-lance testing or s,7ecial tests written spect fically for the application.
A review of the testing performed indicates that the testing of bese circuits was adequate.
The testing was conducted on sixty percent or the leads which were lifted, with no failures noted.
Based on the test results and the fact that all work received double verification and qual-ity control approval, tne licensee has concluded that no further testing is needed.
The items not tested include single wires that were lifted, relugged and relanded on the same terminal, or wires that were lifted and relanded on a new terminal locatio _ _ _ _ _ _ _ _ _ _
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Based on the type of work that was performed, the scope of the testing and the test results, the inspector agreed with the licensee's conclusion.
Recent changes made to the maintenance planning, scheduling, and post-work test procedures will insure that this type of work is more carefully inte-grated into the work schedule such that functional testing is performed by
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scheduled surveillance testing. Post-work test requirements now are spec-ified prior to work in accordance with the licensee's newly developed matrix on post-work testing.
The inspector had no further questions.
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(Closed) Unresolved Item (88-25-01), Failure of Two Residual Heat Removal l
System Valve Yokes. As described in Nf;C inspection reports 50-293/88-25 and 50-293/88-27, the licensee discovered on June 7,1988 that the yoke on the Residual Heat Removal (RHR) System Loop B valve MOV 1001-28B developed a crack 270 degrees around at a weld between the lower yoke section and the motor actuator mounting plate.
(The yoke is composed of a casted steel body welded to a rolled steel operator mounting plate).
Subsequent inspection of the counterpart valve in the A RHR loop, MOV 1001-28A, also identified irdications of cracking in the lower portion of the yoke just below the location of the crack in the MOV 1001-28B yoke.
The li unsee has conducted an extensive root cause analysis of these failures and has undertaken a corrective action program to repair the valves and return them to service.
The licensee formed a task force composed of representatives of the nuclear engineering, operatiens, maintenance and system engineering departments to determine the root cause of the yoke failure and to recom-mend corrective actions.
The investigation conducted by the task force i
revealed that the two main factors which contributed to the valve yoke failures were the def'
" ies in the design and fabrication of the valve (
yoke and excessive mo..
.perator thrust.
i The valve yoke deficiencies included a poor design of the yoke body-to-
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plate weld wnich caused a high stress concentration at the failure point, the low tensile property of the rolled steel MOV mounting plate, and numerous areas of lack of fusion of the weld joining the yoke casting and flange. The design of the 28A and 288 valve yokes is atypical and differs from its original design as a single cast yoke. The yoke original design had been mrdif1ed during initial plant construction so that a larger motor
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operatu could be installed on the valve.
The design of these modified yokes as well as the poor quality of the weld on the yokes contributed to the failure of these yokes.
The other fact;r contributing to the failure of the yokes was the fact f
that the limitorque SMS-5-300 notor operator on the valves delivers much
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more thrust at high torque switch settings than anticipated.
Limitorque testing of the MO-1001-28B motor operator spring pack revealed it to dis-
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play a nonlinear response at high torque switch settings.
The nonlinear response of the torque switeil, combined with the well-lubricated condition of the valve, resulted in the MOV dalivering as much as 200,000 pounds
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..
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more thrust (at the maximum torque switch setting of 4) than originally expected based on the performance data provided by the vendor (i.e.,
545,000 pounds estimated thrust versus 342,342 pounds as stated by the vendor).
In addition, periodic use of the handwheel on the operator by Operations personnel to assist with valve closure added additional stress beyond that delivered oy the MOV.
As a result of the licensee's root cause analysis, the valve yokes for both the RHR 1001-28A & B valves were redesigned and repaired using a new design to reduce the stress concentrations inherent to the original design, incorporate material of a higher tensile strength in the MOV
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mounting plate and utilize welds fabricated under more stringent quality control standards.
In addition, the licensee lowered the torque switch settings on both of the valves (originally set at 3.5 on velve 28A and 4.0 on valve 288) to 2.5 on valve 28A and 1.0 on valve 288 to substantially lower the maximum thrust that the valves would experience during opera-tion, while maintaining the torque switch setting at a level high enough i
to ensure proper valve operation.
Furthermore, the licensee performed MOVATS testing on the valves to ensure that the thrust levels experienced were as predicted.
The inspector reviewed the licensee's preliminary report on the root cause of the yoke failures, a failure analysis report on the yokes prepared by the Massachusetts Institute of Technology for the licensee, Plant Design Change 83-23, Revision 0, dated July 11, 1988 governing the repairs to the valves, the results of the evaluation of the spring pack from MOV-1001-289
conducted for the licensee by the MOV vendor and the MOVATS analysis reports on the 28A and B valves. No problems were noted.
l In addition to the valve yoke failures, the licensee identified cracking in the stellite backseat of the valves.
The cracking was minor on valve
MOV-1001-28A but was extensive on valve 288.
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used on a valve to seal off the process (fluid) side of the bonnet from
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the atmospheric side in event of a packing leak (it is not used during normal operation as a seating area since the valve packing prevents fluid i
leakage to the atmosphere).
The licensee determined the cause of the
backseat damage on valve 28B to be the unwanted contact of the valve with the backseat when stroked to its full open position.
The reason for the
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i contact was found to be the existance of a gap between the stem nut and the lock nut due to the partial outward rotation of the stem nut ' rom its desired position. The gap between tre stem nut and the locknut caused the opening limit switch to allow the motor operator to drive the valve into the backseat, p
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The licensee repaired the stellitt on the 28A and 288 valves using welding procedure P4-A "Overlay Stellite 6" which the inspector reviewed. To pre-vent a recurrence of the backseat damage, improved procedural controls governing the conduct of maintenance on this type of valve were imple-mentyd to ensure that the lock nuts on the valves are properly staked.
The valves were stroke tested and analyzed by the MOVATS testing system to en',ure proper operation of the valves, particularly the limit switches.
The inspector had no further qJestions.
[ Closed) Unresolved Item (88-27-01), Rotation of the Yoke on Core Spray MO-1dOO-4B Valve Body. During a previous inspeccion, the licensee identi-fied that the yoke of the "B" core spray full flow test return line isola-tien valve (MO-1400-48) had rotated out of the correct orientation. Valve MO-1400-4B is a motor operated gate valve. The yoke is held to the valve body by a yoke clamp which was found to be installed incorrectly.
The licensee's investigation determined that an inadequate procedure and msin-tenance personnel error during valve maintenance in August 1987 had con-tributed to the failure. The inspector reviewed the maintenance request package and the quality control (QC) inspection report associated with
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this valve maintenance.
The yoke clamp was installed incorrectly due to
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its synmetrical appearance and lack of match marking during disassembly.
The licensee took immediate action to correctly reinstall the yoke clamp.
l Visual inspection of similar safety-related valves was performed by the licensee to verify proper orientation of yoke clamp positions.
Fifteen valves were inspected and found to be acceptable.
The licensee also revised maintenance procedure 3.M.4-10. "Valve Maintenance", to require match marking and labeling valve components during disassembly and verify-ing the same upon reassembly.
The inspector had no further questions.
Inspector Fe low Items
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(Closed) Inspector Follow Item (85-27-25). Provide Backup Power to the Post-Accident Sample Chiller. The licensee determined that water f rom the fire protection system could be used to cool the samp!e chiller.
The appropriate fittings and adapters were i n s tal l ee..
Procedure No.
5.7.4.1.3,
"Post Accident Sampling System" was revised to include the appropriate steps.
(Closed) Inspector Follow Item (85-27-33), Develop Correction Factors for Non-Isokinetic Sampling of the Reactor Building Vent (RBV) Under Accident Conditions.
These correction factors have been developed.
The licensee also evaluated the impact on sample results for the period betwetn a postulated accident and isolation of the RBV.
Appropriate changes were made to procedure No.
5.7.3.4 "Sampling, Transport, and Analysis of Effluent Iodines and Particulates from the Reactor Building Vent Under Emergency Conditions."
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(Closed) Inspector Follow Item (85-32-23), Establish a Restructured RWP Program. Procedure No. 6.1-022
"Issue, Use and Termination of Radiation Work Permits (RWPs)," Revision 24 was issued in July 1987. This procedure describes an acceptable RWP program.
(Closed) Inspector Follow Item (86-02-04), Licensee to Address Problem
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with ARMS. A review of the calibration of Area Radiation Monitors (ARM)
,
determined that 1) the calibration source strength had significantly decayed; 2) the new fuel storage location was not routinely calibrated; 3) the detector in the TIP room was partially shredded and 4) "as found" readings were not recorded during instrument calibrations. Licensee cor-rective action has been completed as follows:
1) a new full strength calibration source was placed in service in May 1986; 2) procedure No.
6.5-160, "Calibration of the Area Radiation Monitoring System" was revised to require calibration of monitors in the new fuel storage area whenever the area is used; 3) the TIP room detector was relocated to a better geometry in July 1986; 4) procedure no. 6.5-170, "Calibration of Ventila-tion System Radiation Monitors Using ARM Type Sensor / Converters", was revised to require recording of "as found" readings.
Licensee action on this matter is complete and acceptable.
(Closed) Inspector Follow Item (86-02-09), Review ALARA Training Program.
A syllabus was provided of the ALARA training given to managers, super-visors, and tradesmen. This was reviewed and determined to be appropri-ate, A review of attendance record s indicates that many personnel received this training from late 1987 through early 1988.
This matter is closed.
(Closed) Inspector Follow Item (86-06-01). Evaluate the Need to Include Instrument Root Valves on Station Oray h and in Procedures.
During inspection 50-293/88-06 it was i de.it i f i et.
that responsibility for positioning and control of instr.nen; rnot ant isolation valves had not been clearly established.
All valvas were not included in the system operating procedures and all root valves were not shown on the piping and instrument drawings (P&IO).
The licensee has initiated an extensive sys-i
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tem walkdown program to identify vent, drain and instrument root valves.
These walkdowns are continuing and are approximately 50 percent complete.
As the walkdowns are completed procedure and drawing change notices are
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being initiated to incorporate the results.
Those systems remaining have l
been prioritized, with emphasis given to systems inaccessible during plant
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operations.
Responsibility for control of root valves has been clearly assigned to operations.
Instrument isolation valve lineups have been docunented and issued in procedure 8 M.1-33, "Instrument Walkdowns." The Instrument and Controls (I&C) Division will implement procedure 8.M.1-33.
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The licensee has committed to complete the system walkdowns, pror.edure e
revisions, and drawing changes 180 days af ter restart.
Based on the
resources allocated to the ef fort and the progress made to date this goal
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is achievable.
System walkdowns b/ the inspectors indicate that most
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valves have already been incorporated.
The inspector had no further questions, f
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(Closed) Inspector Follow Item (86-06-09), Licensee to Upgrade ALARA Program.
The licensee has implemented the following procedures:
N0P83RC1 ALARA Program 6.10-008 Installation and Removal of Temporary Shielding 6.10-010 ALARA Suggestings 6.10-013 ALARA Job Reviews 6.10-014 ALARA Committee Guidance 6.10-015 ALARA Goals 6.10-0.17 ALARA Procedure Review
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SI-RP.010 ALARA Daily Exposure Review SI-RP.0105 ALARA Reports SI-RP.0205 Engineering Contr:;ls Use Criteria SI-RP.0305 ALARA Plant Design Change Review SI-RP.0315 ALARA In-Process Review SI-RP.0410 ALARA Cost Benefit Analysis SI-RP.0425 ALARA Trending SI-RP.0430 Dose Reduction Data Sheets SI-RP.0900 ALARA Audit These procedures are of good quality and provide a well documented basis for a sound ALARA program. Plann.1g of work is now handled formally by a new department. ALARA reviews are now required as part of work packages.
The use of "A priority" maintenance requests (MR) to circumvent normal work planning and processing has been reduced to a low level.
The "A priority" MR is no longer an ALARA concern.
Action on this item is complete.
(Closed) Inspector Follow Item (86-19-10.1), Provide Formal Instructions for Maintenance Personnel Regarding Decontamination and Storage of Contam-inated Tools.
The licensee issued procedure SI-hT.0401, "Contaminated Tools Control and Storage," in April 1988.
This procedure provides ade-quate guidance for the control of contaminated tools.
(Closed) Inspector Follow Item (86-19-10.3), Licensee to Conduct a Safety Review of New Contamie.:ted Material Storage Areas Per Generic Letter 81-38.
A safety review of the culvert storage a rc a was completed in February 1983.
The results were satisf actory.
Safety reviews of future temporary storage areas will be completed in accordance with procedure no.
OM-5, "Request for Trailers, Tiedown Area, Of fice Space." Action on this item is complete.
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(Closed) Inspector Follow Item (86-37-03), dvaluated the Acceptability of Undocumented Lifted Leads in SurveTTlance Tests.
During inspection 50-293/86-37, the insnector observed that calibration of a high pressure
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coolant injection (HPCI) system discharge flow transmitter was not per-l formed in accordance with the established calibration procedure. Proced-
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ure 8.E.23 indicated that the side cover of the Rosemount transmitter should be removed and the test equipment installed.
Instead the tech-nicians removed the cover from a nearby junction box, lif ted the signal cables and installed the required test equipment.
During the current period the inspector reviewed a sample of calibration procedures and verified that the required technique is clearly delineated.
Calibration of a similar instrument was observed, and practices employed during rou-tine calibrations were discussed with several technicians.
The methods observed and as described by the technicians were consistent with the approved procedures.
The inspector had no further questions.
(Closed) Inspector Follow Item (87-14-01), Review Licensee Evaluation of Accessibility to Upper Orywell Ouring Fuel Movement.
A comprehensive licensee evaluation was issued on July 20, 1937 as a result of potential work hazards identified in General Electric Company SIL #354.
The licen-see study concludes that personnel access to the upper drywell is allow-able but must be administrative 1y controlled and a portable shield must be used in the fue' transfer chute.
This evaluation is acceptable.
(Closed) Inspector Follow Item (87-26-02), Control and Storage of Trans-1ent Equipment. This item was reviewed as part of the Integrated Assess-ment Team Inspection (see inspection report 50-293/88-21 section 3.9.2, Page 92) and is closed based on the results of that inspection.
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3.0 Routine Periodic Inspections J
The inspectors routinely toured the facility during normal and backshift
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hours to assess general plant and equipment conditions, housekeeping, and adherence to fire protection, security and radiological control measures.
Inspections were conducted between ten p.m. and six a.m. on September 8, 1988 for three hours and weekends on September 3,17 and 24,1988 for 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. Ongoing work activities were monitored to verify that they were being conducted in accordance with approved administrative and technical procedures, and that proper communications with the control room staf f had been established, The inspector observed valve, instrument and electrical equipment lineups in the field to ensure that they were consistent with system operability requirements and operating proe.edures.
_
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During tours of the control room the inspectors verified proper staffing, i
access control and operator attentiveness.
Adherence to procedures and
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limiting conditions for operations were esaluated.
The inr9ectors exam-ined equipment lineup and operability, instrument traces ud status of control room annunciators.
Various control room logs and other available licensee documentation were reviewed.
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The inspector observed and reviewed outage, maintenance and problem inves-tigation activities to verify compliance with regulations, procedures, codes and standards.
Involvement of QA/QC, safety tag use, personnel qualifications, fire protection precautions, retest req ui rer.:e n t s, and reportability were assessed.
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The inspector observed tests to verify performance in accordance with approved procedures and LCO's, collection of valid test results, removal and restoration of equipment, and deficiency review and resolution.
i Radiological controls were observed on a routine basis during the report-ing period. Standard industry radiological work practices, conformance to i
radiological control procedures and 10 CFR Part 20 requirements were observed.
Independent surveys of radiological boundaries and random sur-
.
veys of nonradiological points throughout the facility were taken by the I
inspecto,
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Cnecks were made to determine whether security conditions met regulatory requirements, the physical security plan, and approved procedures.
Those checks included security staffing, protected and vital area barriers, personnel identification, access control, badging, and compensatory measures when required.
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j a.
System Alignment Inspection
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Control Room High Efficiency Air Filtration System
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During the period the inspector performed a walkdown of the control i
a room high efficiency air filtration (CRHEAF) system, chysical condi-
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tion end configuration of equipment was observed and compared tt
<
design drawings.
Operating procedures and surveillance test proced-ures were also evaluated.
The system condition, and the technical
,
adequacy of procedures reviewed was generally acceptable.
The I
inspector noted that procedure format was not consistent. The licen-see has initiated an upgrade effort which will address procedure
format and content.
This effort however is not complete. Procedure i
8.7.2.7, "Measure Flow and Pressure Orop Across the Control Room High Efficiency Air Filtration System." is an example of an upgraded pro-
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cedure and represents an increase in procedure clarity and quality.
Technical Specification (TS) requirements are referenced, quantita-l tive acceptance criteria are incorporated and procedure steps are t
i clearly written. During the review however, the inspector identified l
the following discrepancies in other procedures:
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1.
Procedure 2.2.46, Revision 15. "Control Room, Cable Spreading and Computer Room Heating, Ventilation and Air Conditioning System," addresses CRHEAF equipment lineup. This procedure how-ever, does not incorporate positioning of manual dampers.
Re-quired damper positions are also not documented during periodic flow balance testing, and no seal or marking is applied to ensure continued correct positioning.
2.
Technical Specification 4.7.B.2.c requires that the CREAF inlet heaters be operable and capable of an output of at least 14 KW.
Each train is equipped with four heat 19g elements with a total destga outp'Jt of 15.4KW.
Licensee Orocedure 8.7.2.8, Revision 6, "Perform a Functional Test of Humidity Controls and Inlet Heater Capabilities of the Control Room Air Fi1+. ration System,"
calculates the output of each of the four elemelts but specifi-cally states that failure to meet the expected total KW output does not represent a failure to meet TS requf rements.
This is in contradiction with TS 4.7.B.2.c.
.
3.
Procedure 8.7.2.7 establishes an acceptance criteria for the single (one of four) heatir.g element which is controlled by the installed humidity sensor.
These acceptance criteria require that a minimre output of 3.8 KW be achieved at the minimum anti-cipated bur voltage of 420 VAC.
Data collected during testing however, 4, likely to reflect normal bus voltages of 490 to 510 VAC. No orrection factor is contained in the procedure so that calculated results can be directly compared to the acceptance criteria.
4.
Procedure 7.1.30, Revision 13. "HEpA Filter and Charr.oal Filter Performance Test Program," implements required system di-octyl-phthalate, halogenated hydrocarbon, and filter delta pressure testing. However, this procedure does not contain any quantita-tive acceptance criteria.
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5.
Procedure 8.E.47.1, Revi sion 5,
' Control Room /Radwaste Filtra-tion System Instrumentation Calibration / Logic Functional Test,"
,
ensures periodic calibration of system instrumentation and func-tional testing of the logic.
Flow switch FSE-102 senses low flow from the operating CRHEAF system train and automatically starts tse standby train. Step 13 of the procedure performs the instrument calibration, including independent verification of
l proper instrument return to service. Step 14 performs the func-I i
tional test of the automatic low flow start logic.
Step 14.g
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states "Simulate low flow at flow switch FSE-102." The inspec-
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l tor questioned the licensee system engineer regarding the method l
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used to simulate low flow.
The licensee stated that either the i
setpoint on the flow switch is increased until the switch trips
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and starts the standby train, or the instrument piping is dis-connected to cause a low flow signal. The inspector questioned the acceptability of manipulating the instrument setpoint subse-quent to completion of the calibration.
The licensee's Systen Engineering Division is evaluating the Inspec-
tor's concerns to de: ermine if procedure revisions, or test reper-formance is required, In addition, previously completed test results are being evaluated to determine if TS compliance has been main-tained.
(Subsequent to the end of the report period, the licensee indicated that the evaluation was complete and all results had met the TS.)
The inspector also cuestioned if these procedures have undergone the licensee's validation process.
This item will remain unresolved pending review of the licensee's response (UNR 88-31-01).
!
The inspector reviewed the licensee's preparation of the core spray system for operation.
Areas reviewed included core spray system operating procedure 2.2.20, Rev. 33, "Instrument Walkdown Procedure,"
8.M.1-33-1, Rev.1, "procedures," 1.3.4, Rev. 37, and the "Procedure Writer's Guide," 1.3.4-1, Rev. O.
The inspector also performed a walkdown of portions of the core spray system outside the drywell to assess the general physical condition of the system.
Due to better accessibility, system B was chosen, although parts of
core spray system A were also inspected.
The physical condition of the core spray B pump room was adequate.
In particular, the inspec-l tor noted adequate housekeeping (a spill kit was available at the roped off area of the pump room), valve tagging, and overall cleanli-
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ness.
The lubrication level (upper and Inwer pump bearings) was inspected and found to be proper.
Based on this review and inspec-
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tion, the portions of core spray system inspected appear to be ready i
to support the ope ra'oili ty requirement of Technical Specification l
3.5.F.5.
t i
b.
Plant Maintenance and Outage Activities Inspection Report 87-57 noted that while preparing for the primary f
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containment integrated leak rate test (ILRT), the licensee observed l
that several torus temperature and moisture elements wre not func-
tioning properly.
Troubleshooting identified circuit faults at a
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torus electrical penetration assembly.
The licensee removed the penetration assembly prctective cover inside the torus and found that it was filled with water.
The penetration is installed vertically through the top of the torus. On both the inboard and outboard sides (
of the penetration, a metal frame is attached on which 2S terminal f
boards are mounted.
Cables passing through the penetration and
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supplying instrumentation inside the torus are also landed on these terminal boards.
A protective cover is boltod over the frame and i
terminal boards on both sides of the penetration.
Design drawings
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specify that cover joints are to be sealed with silicone tape.
The licensee stated that the prottetive cover had not been properly sealed, allowing water intrusion and buildup. The water caused sig-nificant corrosion of the cable connectors, terminal boards and metal
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frame work.
This corrosion and water buildup resulted in the observed electrical circuit faults.
l To correct the noted deficiency in the torus penetration, the licen-see issued Maintenance Work Requests (MR) 87-50-33 and 34 to replace the terminal strips, seal the upper portion of the penetration assem-bly, and drill weep holvs in the base of the assembly to remove any trapped water which might accumulate.
The penetration assemblies were subsequently retested electrically and found to be acceptable.
The inspector reviewed the Maintenance Work Requests and discussed the corrective actions undertaken by the licensee with the responsi-ble system engineer.
Inspection Report 87-57 questioned the class-ification of the penetration as a
"Q" component (requiring quality assurance controls and oversight) but not as a component requiring environmental qua ',1 fication.
Review of the cables which use this penetration along with discussions with the systemt engineer reveal that this penetration does not contain any cables relutring environ-l mental qualification as specified in 10 CFR 50-49. Thus, the desig-nation of the penetration solely as a
"Q" component is acceptable.
The inspector had no further questions.
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c.
Licensee Actions in Response to the NRC Bulletin 88-05
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NRC Bulletin 88-05, dated May 6,1988 and Supplement I to the Bul-letin, dated June 15, 1988 required that licensees submit information
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regarding materials supplied by Piping Supplied Incorporated (PSI),
i and West Jersey Manufacturing Company (WJM).
It was determined that l
these two companies have supplied potentially nonconforming piping l
materials to the nuclear industry.
Licensees were requested to take actions to assure that any suspect materials comply with ASME Code and ASTM design and material specification requirements.
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The licensee has identified 212 flanges and three caps supplied by PSI and VJM.
Fifty-two flanges were located in the plant, 105 flanges and two caps were located in BEco's warehouse and the balance had not been located as of August 3,1988.
Seventeen flanges ere
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tested insi*u by the licensee using the equotip hardness method. all installed flanges tested were found to be acceptable, except one that l
was slightly outside the acceptance band for screening of initial l
results.
A subsequent engineering evaluation cone,luded that this
f t
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flange was acceptable for its intended application.
Thirty flanges from the warehouse were sent to a certified material test laboratory for destructive chemical and mechanical analysis. Three flanges were
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found Mith slightly less than the allowable ASTM /ASME tensile strength.
A subsequent engineering evaluation indicated that the deviations were not significant and the flannes were acceptable for l
the intended application.
The inspector reviewed the licensee's 120
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day response to the Bulletin, dated September 9,1988 and the above i
mentioned engineering evaluations.
Supplement 2 to the Bulletin was l
1ssued by NRC on August 3,1988, allowing licensees to suspend field measurements, testing, records review and preparation of justifica-
,
tions for continued operation.
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d.
Licensee Actions in Response to NRC Bulletin 88-07
NRC Bulletin 88-07, dated June 15, 1988, was issued to ensure that adequate operating procedures and instrumentation are available, and
that adequate operator training was provided, to prevent the occur-rence of uncontrolled power oscillations during all modes of BWR
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operation.
The Bulletin was issued in response to an event at the
LaSalle Station du*ing which the reactor experienced excessive neu-
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tron flux oscillations following a dual recirculation pump trip from power. The Bulletin required that all licensed reactor operators be
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made aware of the events at the LaSalle station within 15 da 3 of receipt of the Bulletin.
In addition, within 60 days, the licensee
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is required to verify the adequacy of their procedures and operator j
training programs regarding actions to be taken in the event of
uncontrolled power oscillations.
i The licensee's actions taken within 15 days of receipt of the Bulle-tin were reviewed in NRC inspection report 50-293/88-25 and found acceptable.
The licensee's response to the Bulletin was received by I
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I the NRC on September 22, 1988 and reviewed by the inspector.
No
problems were noted.
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l In response to the Bulletin, the licensee prepared and delivered
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training mod 91e 88-0-RO-09-01-10 "power Oscillations in BWRs" to all
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licensed opsrators and shift technical advisors (STAS) performing i
shift duties.
To ensure continued training on the concerns of tha j
Bulletin, the module was also included in the two year requalifica-
tion cycle for licensed operators as well as the RO/SRO Hot License
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and STA initial training modules.
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Licenset operating procedure 2.4.18, Revision 4, "Trip of Both Recir-culation pumps", already directs the reactor operators to scram the
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plant ir, the event both recirculation pumps trip. Thus operation of i
the facility in the natural circulation mode under the power / flow L
conditioe experienced by LaSalle is prohibited.
The licensee's
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response w this Bulletin was timely and their corrective actions
approp-iate.
The inspector had no other questions.
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4.0 Receipt of Licensee Keys l
On September 9,1988, the NRC resihnt inspector took into his possession a key ring snd seven keys that had been anonymously supplied to a repre-
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sentative of CURE, Citizens Urging Responsible Energy.
It was alleged
that the keys belonged to Boston Edison Company and that they would open
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various dwors at the Pilgrim Nuclear Power Station.
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NRC regulations, the Pilgrim Security Plan and the facility Technical
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Specifications require that specific areas and equipment be locked and the
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applicable keys controlled.
The inspector determined that none of the
seven Leys are of a type that has any control requirements under NRC i
requirements. This determination was made by physically comparing each of
the seven keys with security, health physics and operations cepartment I
keys.
In addition, key inventories are conducted on a routine basis and
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records show no missing controlled keys.
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l After the inspector completed the above verification, the licensee was
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j provided with a duplicate set of keys and has identified five of the t
l seven.
Four of the keys are for the onsite medical trailer. While the
keys fit both inside and outside doors to the medical facility, they would l
not allow access to any prescription medication that may be locked within
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I the facility.
The fifth key is a common access control key, which
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includes mar.y areas on site such as battery rooms, and is issued to a i
large number of individuals on site.
The areas accessed by this key are not required by NRC requirements to be locked; therefore, key control is
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unnecessary. Other measures are in place, as specified in the licensee's j
apsreved Sacurity Plan, to provide appropriate security to meet NRC
requirements.
The licensee has made an extensive ef fort to identify the remaining two keys, which are on identical blanks of the type commonly j
used on mobile trailer office spaces, but have found no locks that these
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keys fit.
The licensee has decided not to expend further resources in
attempting to identify the remaining two keys. This decision was based on I
the fact that the keys in question are not controlled keys and on the I
extensive effort already expended.
The inspector had no further
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questions, i
i 5.0 Allegation Regardino Quality Ci rol Inspector Qualifications i
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On July 28, 1938, the NRC resident office received an anonymous allegation
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l that licensee Quality Control (QC)
inspectors performing electrical l
insta:lation and maintenance inspections were not qualified, In response, i
the inspectors observed a sample of electrical maintenance and modifica-i tion items or. going in the field, Particular attention was given to the
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adequacy of QC inspections and the apparent knowledge level of the QC
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inspectors.
Inspectors were familiar with the activities and inspections
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were properly performed.
In addition the qualification and training I
records for licensee inspectors were reviewed to verify that all individ-l uals met applicable requirements.
No discrepancies were noted.
The
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inspector had no further questions.
This allegation is considered closed.
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6.0 Management Meetings
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At periodic Intervals during the course of the inspection period, meetings
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were held with senior facility management to discuss the inspection scope l
and preliminary findings of the resident inspectors. A final exit inter-l view was conducted on October 7,1988.
No written material was given to
the 11 tnsee that was not previously available to the public.
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Attachment I to inspection Report 50-2 JJ/8S-31
Persons Contacted R. Bird, Senior Vice President - Nuclear K. Highfill, Site Director R. Anderson, Operations Manager
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E. Kraft, Station Services Manager A. Morisi, Acting Outage and Planning Manager D. Swanson, Nuclear Engineering Department Manager J. Alexander, Operations Section Manager
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J. Jens, Radiological Section Manage-J. Seery, Technical Section Manager R. Sherry, Maintenance Section Manager
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P. Mastrangelo, Chief Operating Engineer D. Long, Security Division Manager
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W. Clancy, Systems Engineer Division Manager
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F. Wo:ntak, Fire Protection Division Manager l
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