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1 I am responding to concems that you brought to the NRC's attention on May 6,1991. You asserted that the NRC staff had previously been informed of problems in the licensce's in-core instrument pra:tices and has been ineffective in their inspections.
t We have informed the NRC Inspector General of your concems. This issue was a subject of
- 1 inspections 89-05 and 89-13. Copies of the pertinent inspection report pages are enclosed. No l
additional NRC action is planned and we consider this issue closed.
We appreciate you informing us of your concerns and feel that we have been responsive to those Should you have any funher questions, or if I can be of funber assistance in this concerns.
matter, please call me collect at (215) 337-5225 Sincerely,
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t bEdward W[ zinger, Chief Reactor r jects Branch i
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Enclosures:
As stated
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Allegation file, RI-91-A-0092 M. Perkins E. Conner
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ENCLOSURE 1-l i
Excerpt from Millstone 2 Resident inspection 50-336/89-05 (2/11/89 -3/23/89) i 4
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j 18 11.0 Incore Instrument Removal Allecation (RI-88-A-0040)
On February 23, an I&C technician informed the resident inspector of con-cerns identified during participation in a work party to remove incore i
instruments (ICIs) frcm the reactor on February 22.
The ICI job chron-ology is listed below.
The alleger had four separate concerns; each is addressed after the chronology.
2/21/89, 10:30 p.m. ICI removal starts with the first crew.
2/22/89, 3:00 a.m. The electric hoist fails; ICI removal continues without the hoist but with the main crane and the load cell in place.
2/22/89, t.0C a.m. The second ICI crew reports and ICI removal contint.es.
l 2/22/89, 5:00 a.m. The job is interrupted when the Safety Department identifies a concern associated with the I&C workers climbing cown to the Upper Guide Structure (UGS) platform.
2/22/89, 10:00 a.m. A safety meeting is conducted with I&C, Safety, Health Physics, and Millstone 2 management.
i 2/22/89, 1:00 p.m. ICI removal resumes following a full crew briefing and with a new electric hoist and with a portable Area Radiation Monitor ( ARM) installed on the refueling bridge by HP pe-sonnel.
There is no licensed senior y
reactor operator coverage.
(1) The alleger stated that, when he started the ICI removal job at ap-proximately 4:00 a.m. on Wednescay, February 22, an SRO was present.
He stated that, to his recollection, an SRO had always been present during past ICI removal jobs. When the job was restarted after 1:00 p.m. on February 22, the alleger noted that an SRO was not present.
The alleger further questioned whether primary containment integrity i
was required to be maintained during ICI removal, and stated that the personnel airlock was open during ICI removal on the afternoon of i
February 22.
Inspector review found that the licensee had provided SRO coverage during the initial phase of the ICI job, but decided to remove it after the 10:00 a.m. safety meeting. Licensee management concluded that ICI removal was not a " core alteration" and that the initial SRO coverage was a conservative measure. The licensee stated during fol-low-up interviews with the inspector that a similar position was taken in the past, as recently as the 1988 refueling cutage. The licensee deemed this action to be fully consistent with the intent of the Technical Specifications (TS), after determination that movenent of ICI detectors themselves created an insignificant reactivity 3
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l change and the ICI removal process could not create the potential for inacvertent removal of fuel or control assemolies.
The bases for the licensee's determinations were provided in writing at the inspector's recuest and were summarized in a February 25 memorandum from the Unit-2 Reactor Engineer to the Unit 2 Superintendent. The inspector's technical review of the licensee's evaluation identified no safety inadequacies.
P The inspector reviewed shif t operating logs and determined that con-tainment integrity requirements were relaxed sometime on February 22_
and were not met for the remainder of the ICI removal. Containment integrity was relaxed to the extent that the personnel hatch was lef t in the access mode with both inner and outer doors open (to facili-tate personnel movement and to lessen duty cycles on the door oper-ating mechanism).
Shift logs show that integrity was established at 8:37 a.m. on February 21 for installation of the upper guide struc-ture in the reactor.
There is no entry in the shif t log for when containment integrity was relaxed, but it was established again at i
6:30 a.m. on February 25.
Licensee management stated that the de-cision to relax containment integrity followec from the decision that the ICI removal activity was not a core alteration.
NRC regulations in 10 CFR 50.54(m) require a licensed senior reactor operator (SRO) to be present during refueling activities. Millstone Two TS 6.2.2.e requires that an SRO with no concurrent duties be present during core alterations to supervise the activity.
- Likewise, TS 3.9.a specifies requirements for containment integrity during core i
alterations.
TS 1.32 defines core alteration as "
.the movement or manipulation of any component within the reactor pressure vessel with the reactor head removed and fuel in the vessel".
TS bases state that the SRO and containment integrity requirements protect against the adverse consecuences of an accident source term Deing generated during movement of fuel or control assemblies.
The TS definition of " core alteration" results in imposing contain-ment integrity and SRO coverage for a wide range of activities. That is appropriate for activities which can cause significant reactivity changes or fuel damage (e.g., movement of control rods or fuel). But the definition also encompasses activities such as installation of reactor vessel lighting and ICI removal; these activities cannot-cause significant reactivity changes or radiation releases.
Licensee evaluation of ICI removal concluded that the activity was adequately controlled by an approved procedure, that there would be negligible impact on core reactivity, that the ICIs could not in-advertently affect control assemblies due to design of the UGS and the ICI plate / thimble tube, and that the refueling boron concentra-tion assured that the core would remain subtritical even without the control roos inserted.
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P Inspector review concluded that ICI removal posed much less risk than activities which could cause significant changes to core reactivity or core damage. However, ICI removal is a radiation hazard if the ICI tips became unshielded; the activity requires experienced per-sonnel, and apprcoriate procedures and supervision.
Other than the f ailure to recognize ICI removal as a core alteration, no inadecuacies we.e identified in the licensee technical evaluations and safety conclusions for this specific instance. After review of this matter with NRC management, the inspector informed the Unit 2 Superintendent in a meeting on February 24 that the Technical Speci-fication should be implemented literally until the definition could be revised by the amendment process.
The inspector noted the ICI i
removal activities had already been completed by February 24.
The licensee acknowledged the inspector's comments, stated that this ap-i proach would be taken for subseouent in vessel work that meets the literal TS definition of a core alteration, and expressed the intent to recuest a Technical Specification change.
NRC review concluced that ICI removal meets the TS definition of a i
core alteration. The associated f ailure to provide SRO coverage and maintain containment integrity in accordance with Technical Speci-fications 6.2.2.e and 3.9.4 is unresolved pending resolution of the TS cnange planned by the licensee (UNR 50-336/89-05-07).
(2) The alleger said that no pre-job briefing was conducted. The inspec-tors reviewed the job anc questioned the HP and I&C departments and
' arned that the HP department conducted pre-job briefings and that ea off going I&C technician conducted on station turnovers and con-i firmed that each individual understood his responsibilities after turnover. The alleger confirmed that these briefings did in fact occur, but he was concerned that a pre-job group briefing was not conducted by the I&C department.
The inspector confirmed that a pre-job group briefing was conducted by the I&C supervisor during the evening of February 21 for the first ICI removal crew. The alleger said that the members c' the second crew would have benefitted from a group briefing and indicated that the contractor personnel were un-familiar with the job and the radiation hazards involved. He also stated that the NNECo personnel were uncomfortable with the contrac-tor personnel, pointing to a specific example where a NNEco load director would not take direction f rom a contractor and the alleger himself was repositioned to communicate with the load director.
I The inspector spoke with the members from the first and second crews and concluded that a pre-job brief at 4:00 a.m. on Februa y 22 would not have substantially improved the job.
Although not required by the ICI removal procedure, the first c.ew was given a briefing on the procedures and individual duties prior to the start of the evolution.
That same supervisor concluded that an on-station turnover was suf-ficient to assure the 4:00 a.m. relief crew was adequately familiar j
with the task and their individual responsibilities. The inspector 1
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noted further that there was an additional-full crew briefing con-ducted during the 10:00 a.m. meeting on February 22 for all I&C per-(
sonnel that could subsequently be affiliated with the job.
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The alleger also stated that the lack of a pre-job briefing caused two or three crew members to give directions to the.rane operator.
I When interviewed by the inspector, the load director stated that only i
he could talk to the crane operator by using headset communications, i
due to the distances involved.
Further, the load director stated he was not affected by the " speakers" as he stood next to the load cell 3
spotter, watched the load cell himself and would have directed the i
crane operator to stop the upward motion if the 250 pound load limit l
was approached.
The load director also stated that he did not want to use a contractor as a load cell spotter because that would be in conflict with (uncocumented) routine maintenance practices on the control of loads.
The load director stated that he had confidence in the capability of contractor personnel to adequately perform the i
work.
L While all second crew workers contacted appeared somewhat uncomfort-able with the lack of supervision present, there were no adverse con-sequences. The inspector concluded that the presence of I&C supervi-sion during ICI removal would have been beneficial to coordination among crew members. However, while no pre-job group briefing was
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conducted for the second crew, the inspector concluded that pre-job briefing would not have substantially improved the conduct of ICI removal in this case, and was not required.
(3) The alleger stated that he encountered the first line I&C supervisor f
for the first shift of ICI removal prior to shift turnover.
The I
supervisor stated that the electric winch had failed and that he chose to delete the winch and perform _the pulls with the main hoist alone.
The alleger questioned this practice and the supervisor 7'
stated that he considered it an equivalent if not better method.
The alleger brought up two concerns o's this issue:
r The electric hoist allows the person who is observing the load I
cell to stop the lift if load reaches the 250 pound limit speci-t fied in the procedure, IC 2419A. The person watching the load cell has to notify the load director, who notifies the crane operator. The alleger feels that lifting with the main hoist
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alone is less safe because communication is more difficult and indirect.
t The alleger stated that, because the electric winch was not r
used, the procedure was violated and an interim procedure change i
had not been prepared.
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A 22 The job began by using both the electric hoist and the polar crane
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auxiliary hoist as specified by procedure step 5.2.I7.
When the electric hoist f ailed, the I&C job supervisor decided to continue with the ICI pulls with the main hoist alone until an alternate hoist could be obtained.
The procedure requires that the electric hoist be used to pull the ICI as f ar as the hoist chain will allow while ob-serving the spring scale and guide tube.
Step 5.2.I8 allows use of the auxiliary crane hoist to continue to pull the ICI clear of the guide tube.
The procedure does not specify the electric hoist chain length.
In the past, the licensee has used electric hoists with either 20 feet or 40 feet of chain length; a 40 foot hoist was used during the first part of the job on February 22.
The polar crane hoist would be used depending on which electric hoist was available and the indivicual ICI length.
The longest ICI must be pulled at l
least 30 feet to clear the guide tube.
A total of two to six ICIs were removed without the electric hoist between 3:00 and 5:00 a.m. on February 22.
The inspector reviewed the configuration without the electric hoist and concluded that it is a safe way to conduct ICI pulls. While the pathway to stopping a pull is less direct, it is acceptable because the load cell cbserver stands next to the load director, who uses a i
head set to maintain constant communication with the crane operator.
The safety of the method is f urther supported by the fact that the crane speed used, SI, is slower than the electric winch speed.
In addition, IC 2419A specifically allows ICI pulls with the crane when the electric winch runs out of chain length. The inspector concluded j
that any additional time delay in stopping an ICI pull had a negli-t gible effect.
i Af ter discussic.s with several crew members, the inspector concluded that they had c1ffering views on the best way to conduct ICI removal (that is, with or without the winch).
The alleger also stated that modifications to the ICI removal equipment (such as the installation of a remote load cell readout for the crane operator and mechanical interlocks to prevent pulling the rhodium detectors out of the water) would improve the safety of the job. The inspector concluded that the current method is acceptable and that the job was conducteo safely on February 22.
As for changing the ICI removal method without making an interim change to the procedure, licensee management and the I&C first line supervisor stated that, because the method was equivalent if not better than the original method and the intent of the procedure was met, it was within the authority of the supervisor to continue with the revised method without changing the procedure. The licensee in-formed the inspector of a previously established licensee position on the authority of supervisors and test directors to proceed with jobs if certain procedure steps do not apply because system conditions are off-normal. The supervisors may then determine that the intended 3
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_m-23 system condition is satisfactory and that no procedure change is necessary prior to continuing with subsequent steps.
This position was developed in proposed Revision 44 to ACP 3.02.
The 50RC (Station-4 Operations-Review Committee) approvea the proposed changes in 50RC Meeting 88-43 on December 20, 1988 and made them ef fective for im-i' plementation on March 10, 1989. Within the above framework, ACP Rev 44 Section 6.6.2 establishes -limited criteria which, if met, allow certain procedure deviations that do not require a procedure change.
Review of ACP 3.02, " Station Procedures and Forms," and discussions with the licir.ne addressed how the administrative procedure provides criteria for use by plant personnel en compliance with written pro-cedures. As a station administrative procedure, ACP 3.02 applies to all three Millstone units and establishes procedures as the primary job aid for performing work at Millstone.
The provisions in ACp 3.02 include: use of temporary or substitute instrumentation; definition i
cf when a supoort system is considered available; guidance on how to l
resolve conflicts in approved procedures; conditions under which de-l viations from system valve lineups and checklists are allowable; per-forming steps concurrently; not completing a procedure; when to make i
formal changes (i.e. either pre-approved by PORC or SRO approval with follow up PORC review); and actions allowed in the event of emergency i
conditions.
The following excerpts from ACP 3.02 delineate management's expec-l tations on procecure adherence.
procedures support maintenance, modification, surveillance and operation by providing detailed in-'
i formation to the user that should contribute to job efficiency, per-sonnel safety, error minimization and radiation exposure reduction.
Procedural detail must be suf ficient for a knowledgeable user to per-f orm the evolution correctly. A successful tast is.he sum of moti" l
vated and qualified personnel in addition to the tools at their dis-posal. Training and experience directly contribute to qualifications while procedures complement the tools.
The expectation on the use of procedures is for workers to review the procedure prior to start of work; review all steps, notes and cautions prior to start of work, i
and if not understood, obtain clarification from a qualified indi-i vidual; follow the procedures explicitly in the order written unless the procedure allows exception or the provisions defined in ACP 3.02 section 6.6.2 apply; and correct any procedural deficien:ies upon identification, including stoppage of work activities to formally change the procedures if necessary. The need to formally change the procedure is defined to exist when the procedure will not work as written and the suggested changes will be permanent.
Upon review of proposed Revision 44 to ACP 3.02, the inspectors iden-tified several open issues whit.h include: the need to clarify the criteria, the adequacy of the criteria in view of TS 6.8.1, speci-1 fication of the minimum level of qualification or supervision needed to implement such procedure changes, and the need to document the l
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24 bases for such changes.
The inspector presented these concerns daring a March 9 meeting with the Station Superintendent. The lic-ensee deferred Revisic,n 44 implenntation pending further review and clarification of the intended changes.
In summary, NRC inspection of the alleger's two concerns related to implementation of IC 2419A concluded: (i) actions to pull ICIs with-out an electric winch were a deviation from the method prescribed by IC 2419A, but met the intent of the procedure and constituted a safe, acceptable method of removing ICIs; and (ii) the job supervisors acted in accordance with management expectations as contained in existing and/or pending administration procedures. Adequacy of and licensee actions to clarify and revise ACP 3.02 will be reviewed on a subsequent routine inspection (UNR 89-05-08).
(4) The alleger stated the #2 Safety injection Tank (SIT) area radiation monitor (ARM 7891) was out of service for calibration.
He noted that the HP (health pnysics) department was not aware of its unavailability and indicated that there was a lack of coordination between depirt-ments.
The alleger said that he questioned the absence of an ARM and that a portable ARM was installed af ter he raised the concern to an HP technician during the 5:00 a.m. to 1:00 p.m. break.
The inspector investigated the HP coverage for the job and concluded that the protection af forded by the alarming dositecs (digital alarm-ing dosimeters) affixed to each worker and an HP technician's use of a teletector was adequate to ensure personnel safety. The ARM would only have provided more defense in depth.
The inspector spoke with the HP technician who confirmed that the alleger informed hb of the absent ARM, and stated that the decision to place a portable ARM had been reached at a prior HP planning meeting and that HP was late in placing the ARM on the refueiing bridge.
The alleger also stated that he thought that the ARM was supposed to undergo a setpoint change for refueling conditions and that he sus-pected that this had not t,een done. The inspector reviewed the set-point change issue with the Operations and I&C departments, who stated that the ARM does not undergo a setpoint change for refueling operations.
The inspector questioned the licensee's delay in ;alibrating the #2 SIT ARM, which was removed from service on Febrcary 28. The inspec-tor noted that it is the closest normally-ava dable ARM. The I&C technician responsible for calibration of the ARM stated that there were numerous detector wiring problems that delayed its return to service until March 1.,1989.
The licensee stated that ARM unavail-ability will be added to the outage critique list and they are consi-dering the purchase of a replacement ARM which would be calibrated i
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t 25 prior to the next outage and placed in service so that the normally-f available ARM coverage is maintained. The inspector will follow the licensee's resolution of this issue during routine inspections.
12.0 Committee Activities (71707) r The inspector attended Plant Operations Review Committee (PORC) meetings 2-89-22, 2-89-28, 2-89-29, 2-89-30, 2-89-35, 2-89-46, 2-89-48, 2-89-54, and 2-E9-56 on 2/10, 2/15, 2/15, 2/16, 2/20, 2/27, 3/1, 3/10, and 3/13.
Committee administrative requirements were met for the meetings, and the committee discharged its functions in accordance with regulatory require-ments.
The inspector observed a thorough discussion of matters before the PORC and a good regard for safety.
No inadequacies were identified.
Tt.e inspector also attended meetings of the PORC (2-89-37) and the Nuclear Review Board (2-89-3) on 2/22, held to review a proposed change to Tech-nical Specification 3.9.3.2 requirements on spent fuel pool cooling. The reviews by both committees were thorough and technically sound.
The pro-posed technical specification request was subsequently not submitted due to a change in p' ant conditions.
No inadequacies were identified.
13.0 Licensee Event Reoort (LER) Review (92700)
Licensee event recorts s;bmitted during the period were reviewed to assess LER accuracy, the adequacy of corrective actions and compliance with 10 CFR 73 reporting reauirements, and to determine if there were any generic implications or if further information was required.
The LERs reviewed were:
LER 89-001-00 " Fire Barrier Penetration Seals Inoperable" LER 89-002-00 " Main Steam Safety Valve Setpoint Drif t Uncovered During As-Found Simmer Test" LER 89-003-00 " Combined Leakage Rate Exceeded" No unacceptable conditions were identified.
LER 89-001 review is ad-f dressed in Detail 12.1, and LER 89-002 is considered in Inspection Report 50-336/89-03.
i 13.1 LER 89-001-00, " Fire Barrier Penetration Seals Inoperable" On February 2, at approximately 5:00 p.m., the licensee determined that two fire barrier cable penetration seals were inoperable. -The cable penetrations were identified as numbers 108 and 109, located between the mai.n cable vault and east electrical penetration room.
The inspector reviewed the following documents in follow-up of LER 89-001-00:
ENCLOSURE 2 Excerpt from Millstone 2 Resident inspection 50-336/89-13
65 Since the filters are routinely changed by health physics technicians, the inspector inquired f rom the Unit 2 Health Physics Supervisor if this situation was known to the Health Physics (HP) department. He stated that the HP group had generated a Trouble Report to I&C approximately one year earlier. However, the inspector -as unable to obtain a copy of the Trouble Report. A work order is currently on-hold pending receipt of the parts from the vendor.
Conclusion l[
This allegation is substantiated in part; however, a nonconformance report was written to address that ceficiency as required. More importantly, current 1&C personnel had no knowledge of an outstanding Trouble Report on this issue from the HP Department. Upon redisccvery of the problem by the alleger. A nurchase order and NCR were issued although the timeliness of I&C supen'ss~y approval of the NCR (almost 6 weeks) could have been improved. The screen functions to hold the filter in place within the samole flow and a modified screen placed in the filter holder mechanism would not ccmpromise the monitors intenced safety function.
A.9, 8.6.2 Incore Instrumentation Allecatien This allegation acdressed concerns that the removal of incore instruments (i.e., incore neutron cetectors) during the February 1989 outage was poorly controlled and potentially unsafe. Among the concerns was that the procedure was unsafe; there was no pre-jcb briefing; the procedure was not used; the procedure was changed without an approved procedure change or safety review; there was no licensed senior reactor operator present as required by technical specifications for core alterations; there was inadequate health physics coverage; the previous shif t cn February 22, 1989 was poorly supervised.
Note:
This allegation is related to B.6.2 which covers the same topic; hence B.6.2 is incorporated in this discussion.
Discussion The above allegation was previously given to the NRC resident inspectors office and was addressed and closed per WRC Inspection Report 50-336/89-05, issued May 4, 1989. All of the above allegations were addressed in that report and it is not the intent of this inspection to readdress each issue. This inspection does perform an independent safety review of the ICI removal operation and the adequacy of the health physics coverage.
It should be noted that during the original NRC inspection, ICI removal was in progress.
During this inspection, the reactor was operating and the refueling floor was inaccessible.
The inspector discussed the ICI removal job with the personnel directly involved, an I&C supervisor and the I&C engineer. However, the actual evolution could not be witnessed as the plant was operating during this inspection. The inspector also reviewed the following procedures.
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66 IC 2419A, ICI Replacement - Installation Procedure, Revisions 0, 8 and 9 Change 1 OP 2352, Polar, Pedestal, Cask, and Turbine Building Crane Operation, Revision 4, Change 4 IC 2442G, Local Area Radiation Monitors Calibration, Model GA2 TMO, Revision 2 OP 2383C, Radiation Monitor Alarm Setpoint Control, Revision 1 HP 904/2904/3904D, Calibration of Fixed Monitors, Revision 9 No.1370-CCE-GL60-4, Revision 00, Guidance For Removal and Disposal of the incore Instrument Assemblies For San Onofre Nuclear Generating Station Unit 2&3 Discussions with various personnel appear to validate the e ployees complaint that the ICI removal shif t f rom 8 p.m., February 21 to 4 a.m., February 22, 1989, lacked proper supervision and manacement.
These problems centinued into the next shift beginning at 4 a.m.
The ICI removal job was stopped at 5 a.m. for other safety problems. Ultimately a safety meeting was held at 10 a.m. to straighten out coordination between organizations, improve safety and gain control of the job.
The inspector reviewed the employees main concern that removal of the ICI's using only the polar crane rather than an electric winch as required by the procedure was an unsafe practice and increased potential radiation exposure to personnel.
The electric winch had failed on the 8 p.m. - 4 a.m. shift and was removed. A replacement winch, although available at the site, was not used.
The ICI removal continued using the polar crane only. Although procedure IC 2419A required the use of an electric hoist, no procedure change was obtained. ACP-QA-3.02, station procedures and former Revision 43, Paragraph 4.8 states in part, "An intent change is one which includes a change in the basic method of the procedure, a change which could endanger personnel... Examples [ include]...
A significant addition or deletion of procedure steps which is not consistent with the original " applicability of the procedure". The deletion of the winch and use of the polar crane appears to be clearly an intent change requiring management review and a safety evaluation. Although this issue was left un-resolved in NRC inspection 50-336/89-05, further review during this inspection indicates that a change should have been obtained.
Failure to obtain an approved change is contrary to Technical Specification 6.8.2 and is considered another example of f ailure to follow procedures and is discussed in paragraph A.6.4.
Unresolved item 336/89-05 remains open because of other issues which require resolution.
Although a change should have been obtained, the operation was safe.
The licensee had changed to using an electric winch for ease of doing the job considerations.
The inspector reviewed Revision 0 to IC 2419A and noted that a hand chain pull I
was used to initially move the ICI's and then the polar crane was used to complete the lift of the ICIs. A review of a San Onofre (another Ccmbustion Engineering) l
67 procedure for ICI removal also uses the polar crane for ICI removal (the ICI cables at San Onofre are solid rather than wire). The use of the polar crane, if properly centrolled, is a safe and acceptable method of ICI removal.
The employee was also concerned with the potential loss of communications to the polar trane Operator as af fecting safety of ICI removal. The inspector reviewed crane operating procedure OP 2352 and observed the following statements:
"6.8 any time the crane operator is unsure of the operation called for by the signalman, or the signal is unclear, the operator should stop all motion until the communication problem is resolved."
"7.1.2. CAUTION:
If racio communications are lost, all movement of the polar crane 1s to be stopped until communications are restored, or it is determinec by upper management that the operation in progress can be completec safely by use of hand sionals.
The above caut:cns to the polar crane operator and the use of spotters from the refuel bricge assure safe operation.
The improper use of the procedure IC 2410A could not be confirmed. According to the employee, the procedure was at the job site.
Conceding that the 8 p.ra. -
4 a.m. was not dene preperly, there appears to have been better control of job after the 10 a.m safety meeting with management.
The employee stated that there was improper health physics oversight of the job and that an area radiation monitor was out of service.
It was confirmed that the area radiation monitor was out of service. However, health physics personnel were not dependent on this equipmer.t.
Available at the job site was a portable Teletector radiation monitor plus alarming personal dosimetry devices for l
personnel working the ICI job. This was adequate for the job. However, Change 1 l
to Revision 9 to procedure IC 2419A has added the following prerequisite "3.3.1 area radiatien monitors RM-7890 anc 7891 are in operation or a portable monitor is provided by the Health Physics Department".
The setting of area alarms is at the discretion of operators as per procedures OP 23838 and 2383C. Area radiation monitors RM 7890, containment personnel access and P,M 7891, containment refuel machine are listed in these procedures as being set by operations to suit background conditions as long as MPC limits are not exceeded. Figure 6.5 of procedure HP 904/2904/3904 D lists the alarm setpoints for RM's 7890 and 7891 as 225 mR/ hour operating and 100 mR/ hour shutdown. However, this procedure is for guidance only and procedures OP 2382 B and C take precedence.
Conclusions The various allegations concerning the ICI removal job on February 21 and 22 were, in part, substantiated. The chief allegation, from two employees, that the removal of the ICI detectors using only the polar crane was an inherently
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68 unsafe operation is unsubstantiated.
There is ample precedent that this operation, reasonably controlled, is safe.
The use of an electric winch attached to tN polar crane is a better controlled and more convenient operation but that does not mean use of the polar crane alone is unsafe.
The employees' view that a procedure change and safety review should have been obtained is substantiated. The deletion of the electric winch from use is clearly a significant change to the procedure.
It also is a significant change from the intent of the procedure. A formal change with PORC review and unit superintendent approval should have been obtained.
That proceaure IC 2419A was not used was not substantiated.
The procedure was at the job site. Procedure use does not necessarily mean a procedure has to be in hand, being read step oy step but it should be followed. Conceding that the 8 p.m. - 4 a.m. snif t on February 21-22, 1989 did not function properly, there is no evidence of general misuse of this procedure.
The allegation that the health physic: centrols were inadequate, was unsub-stantiated.
Health physics has the latitude of setting their own controls.
j They co not depend on plant installed area radiation monitors nor do they have control of the alarm setpoints which L.e set by operations.
Procedure HP 2904 which prcvides radiation monitor setpoint guidance does not apply.
The allegations concerning no pre-job briefing and the absence of a Senior Reactor Operator f rom shif t during ICI removal were addressed in 50-336/89-05 and will not be addressed f urther in inis report.
A.10 Radiation Monitor Low Setooint Alarm Allegation This allegation refers to RM-9116, " Aerated Waste Liquid Discharge Monitor "
"This is a PIOP's, NMC rad conitor - it has a micro processor in its base. The issue is that the alarm set point in terms of the high alarm set point can be changed, and is changed on a normal everyday basis by the plant equipment operator af ter chemistry analyzes the sample. He (operations) inadvertently changed the low alarm and didn't know that he did it. I don't think that is an unsafe con-I dition. So they gave me an automated work order (AVO) and said that the fail alarm was in all the time. I accessed the computer and it told me exactly the i
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same number for the high and low, so I recognized what I think he did. What I did find in the interim was that there wasn't any place that I could go to find out what the low alarm should be. Since the PIOPs monitor was installed three or four years ago, nobody has ever derived a low set point alarm setting in engineering. We took the numbers that were in our procedure - 1E3. The question is in the safety review, the PDCR safety reviews, and the retests that all were done on that rad monitor, why didn't we never derive a low alarm set point?"
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.s AUG 1 ? 1991 j
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I am responding to concerns that you brought to the NRC's attention on May 6,1991. ~ You asserted that the NRC staff had previously been informed of problems in the licensee's in-core instrument practices and has been ineffective in their inspections.
i We have informed the NRC Inspector General of your concerns. This issue was a subject of
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inspections 89-05 and 89-13. Copies of the pertinent inspection repon pages are enclosed. No j
additional NRC action is planned and we consider this issue closed.
We appreciate you informing us of your concerns and feel that we have been responsive to those Should you have any further questions, or ifI can be of further assistance in this concerns.
i matter, please call me collect at (215) 337-5225 t
.t Sincerelv Origirial Tigned By:
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/'bEdward Wenzinger, Chief i
Reactor Projects Branch 4 1
Enclosures:
As stated i
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Allegation file,-
RI-91-A-0092 j
M. Perkins i
E. Conner l
W. Raymond i
EG&G Idaho Representative (CARDONE)
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P02 P(-W{91! 13:05 NRC PilLLSTONE 0FFICE 1
l ALLEGATION RECEIFT REPORT I
Date/ Time NW l 1%l !!2:l'IcM.
Allegation No. &91 dd9 3'-
l (leave blank) l l
Received:
~
Address-f uare City / State / Zip:,
.i Phone:
Yes ~ No ~/
l Confidentiality:
Yes No Was it requested?
No '~~~
Yes _
I vas it initially granted?Vas it fins 11y granted by the allegation pa Oces a ccnfidentiality agreer.ent need to be sent Yes No ~
No ]
to alleger?
Yes Has a confidentiality agreement been sigr,ed?Meco No Yes Fosition/
Title:
fe UNE00 i
NitLhTot)E2 Docket No.: 60-336 Faci)ity: _
6). IMPrCW j
(Allegation Sumary (brief description of concern (s):r O thchafSufWd[Qr.(C,
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l \\ ~ l l f j Number of Concerns: __ 9 7 N A 616 N O E b T' l Empicyee Receiving Allegation: (first two initials ana last name) i l'fg y5 Type of Regulated Activity (a _/ Reactor d _ 3afeguards'
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i - SI\\ .e. Other: (5pecify) l Vender ~~ ~ Materials
- e jg j %h Materials License No. (if applicable): _
f j i; C= 7gg _ e) Emergency Preparedness Functional Ares (s)':,/,_(a) Operations Onsite Health and Safety Offsite Health and Safety-Construction ~~ ~ e s M _ s-Safeguards ~~ Other: 5s#e Transportation I ( (NRC Region 1 Fom 107 -Revised 10/89) / um. ws exw, h,,4%
f AY 07 '91 13:07 NRC MILLSTONE OFFICE t-ve SA-34 ~ S ADV ANCE SYSTEM _ advance system is driven by a constant f tor through rapidly interchangeable gears withAccordi inuous f these gears provides a speed ratio of 4 to 1.
- ,r 2 to 1.
t speed capstan provides an advance speed "O" ring driven take-up vides a constant tension take-up for the The advance speeds are normally one '.1/4 inch per hour; however, changing the con-sper. eed drive motor provides the option of chancing ,o speeds. PER SUPPLY in length. The 'or filter paper is 125 feet 't hours per roll depends upon the filter speed. eed of 1 inch per hour the system uses 2 feet per ?accordingly, the filter paper roll will last ~._ __,,, N.. N s , ALE CHECK MECHANISM 'N N cale check mechanism comprises a remotely con-1: i 137 source ,' motor-driven hermetically sealed Ces um Since ing added activity in the counting chamber. scale ch jty is present on the paper. final activity counted >c the arithmetic sum of the counts from the paper / operation of the system is he source. Correct by seeing that when the upscale check source is uced into the chamber the counting level will be st the level of the check source. j ~' 2.0 - Thr.0MI ur OTERATI5d ~. I F tCT AND BACKGROUND inch thickness of lead provides a 90 to. LOO fold The background of f tion for a Cobalt 60 field. -2B detector mounted in the SA-34 shield is 20 c/M t an elevated external field. 4 h. ai o.
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e.o4 SP 2404Ag-Page 5 Rev.2 fA f=1~ m 6.2. Count Rate Module Electronics Functicnal Checks l BJOTE: I&C bstruction 1.06 gives guidance on interpolating logarithmic scales. Any questions should be referred to 1&C Supervisor for resolution. 6.2.1. At the Local Alarm Panel, BYPASS the field alarm horn for R51-9095 by LNSERTING the Alarm flypass Key and TURNING it to the BYPASS Positien. 62.2. PLACE the CP61 Test /OP switch in the test position and RECORD the As Found electronic test reading fer CRM module RM 9095 meter on 1&C Form 2404AG-1. 6.2.3. ALLOW the Test:OP switch to Spring Return to the OPERATE position. RECORD the CPal reading in the Background column of I&C Fcrm 2404AG-1.- [~ NOTE: ( The Acceptance Criteria is that the Urscale Check is > ( Brkground, not that a specific incrce3.0,c_ctirt.,_.__ ~ 6.2.4. PLACE the Upscale Check switch in the 1N position and VERIFY the white Check Source Drive light comes on and the CRM meter reads increased counts. VEFJFY the v;hite Check Source Drive light goes out when the CPSI meter reading stabilizes. RECORD the CRM reading in the Upscale Check column of 1&C Form 2404AG-1. 6.2.5 RELEASE the upscale check switch to the out position and VERIFY the white Check Soerce Drive fight comes on and the CPOI meter reads decreased counts. VERIFY the white Check Source Drive light goes out when the CRM meter reading stabilizes. 6.2.6. CONNECT the RC 14 test current source to the CRM module Test Jack and ADJUST the Current Trip Test Potentiometer (CTTP) for a value between the HIGH and FAIL alarms. 6.2.7. RESET the CRM module alarms. jr:fx.asa in :n,3,gc;g m gg3 !D 2:=d::c riith i e frn20- 4 in cr,.uW s SCl. EXETUp%ns _ Y_~7 ^ F01A f 2 -/_[ F._ ~ ~ ~ - ,}}