IR 05000275/1993011
| ML20059M827 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 08/24/1993 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| References | |
| OLA-2-I-MFP-026, OLA-2-I-MFP-26, NUDOCS 9311190363 | |
| Download: ML20059M827 (21) | |
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UNITED STATES
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9 j#muc%,t NUCLEAR REGULATORY COMMISSION a j )p'E REGION v '9 PDC: 28 P6 20 t 8 1450 MARIA LANE si J
\ i 8 WALNUT CREEK CAUFORNIA 94596 h DC)@^npfugRY Officist Exh. No Dettet No MAY.1. 319N is. rah m t A R / M C D.l_6 LN M# 4 [; i 8"
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9 g Dockets 50-275, 50-323
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Licenses DPR-80 and DPR-82 - """
EA 93-107 lea Y - .
Pacific Gas and Electric Company 77 Beale Street, Room 1451 3hb Som wan ss - g San Francisco, California 94106 m,
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Attention: G. M. Rueger Senior Vice President and General Manager l Nuclear Power Generation .
i NOTICE OF VI0t.ATION j SUBJECT:
NRC INSPECTION REPORT 50-275/93-11 AND 50-323/93-11
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During the period of April 5 - 9, 1993, Mr. L. Coblentz and Mr. K. Brewer of this office conducted a routine onsite inspection of activities authorized for your Diablo Canynn Power Plant. At the conclusion of the onsite inspection, 1 Mr. Coblentz and Mr. Brewer discussed our findings with members of your staff $
identified in the enclosed report. Additional inspection was conducted in the .-
Region V office from April 12 - 28, 1993, and these additional findings were ;
- discussed via teleconference on April 29 and 30,1993, with members of your
$ staff identified in the enclosed repor Areas examined during this inspection are described in the enclosed repor Within these areas, the inspection consisted of selective examinations of .
procedures and representative records, interviews with personnel, and g observations of activities in progress. The purpose of.the inspection was to '>
determine whether activities authL,'ized by the licenses were conducted safely
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and in accordance with NRC requirement Based on the results of this inspection, certain of your activities appeared to be in violation of NRC requirements, as specified in the enclosed Notica of i Violation (Notice). In four instances from October 1992 to March 1993, you !
identified that individuals had enterei posted l1igh radiation areas without I meeting the radiological controls required by your~ technical specification ;i t
Because these instances were self-identified, they would normally meet the ..
criteria of Section VII.B.2 of the NRC 2nforcement' Policy for non-cited }$
violations; however, the multiple instances indicate that your corrective .id measures to prevent additional occurrences have not been fully effectiv p You are required to respond to this letter and should follow the instructions specified in the enclosed Notice when preparing your response. In your hl
response, you should document the specific actions taken and any additional !
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actions you plan to prevent recurrence. After reviewing your response to this -
Notice, including your proposed corrective actions and the results of future :
- i inspections, the NRC will determine whether further NRC enforcement action is I
9311190363 930824 PDR ADOCK.05000275 Q PDR
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NOTICE OF VIOLATION Pacific Gas & Electric Company Dockets 50-275 and 50-323 Diablo Canyon Power Plant Licenses DPR-80 and DPR-82 During an NRC inspection conducted on April 5 - 28, 1993, a violation of NRC requirements was identified. In accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2, Appendix C, the violation is listed below:
Technical Specification (TS) 6.12.1 requires that high radiation areas in
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which the intensity of accessible whole body radiation is greater than 100 i.nilirem per hour (mrem /hr) but less than or equal to 1000 millirem per hour (mrem /hr) at 18 inches shall be barricaded and conspicuously posted as a high radiation area. Any individual or individuals permitted to enter such an area must have either: (1) a radiation dose rate survey meter, or (2) an alarming dosimeter and prior knowledge of the area dose rates, or (3) accompaniment by an individual qualified in RP procedures h who carries a survey meter, provides positive control over activities in the area, and performs periodic radiation survey Contrary to the above:
. On October 2,1992, during the Unit 1 1R5 outage, an individual entered the refueling bridge posted high radiation area in the Unit 1 Containment Building without having either a survey meter, an alarming dosimeter, or accompaniment by an individual qualified in RP procedure . On October 26, 1992, during the Unit 1 1R5 outage, two individuals entered a posted high radiation area on the 140-foot elevation of
's the Unit 1 Containment Building without having either a survey meter, alarming dosimeters, or accompaniment by an individual qualified in RP procedure .- 3P" On March 15,1994 during the Unit 2 2G5 outage, three indchiduals entered a posted high radiation area adjacent to reactor coolant pump 2-2 in the Unit 2 Containment Building without having either a survey meter, alarming dosimeters, or accompaniment by an individual qualified in RP procedure . On March 22, 1993, during the Unit 2 2R5 outage, two individuals entered the posted high radiation area inside the pressurizer shed >
in the Unit 2 Containment Building without having either a survey meter, alarming dosimeters, or accompaniment by an individual qualified in RP procedure This is a Severity Level IV violation (Supplement IV).
Pursuant to the provisions of 10 CFR 2.201, Pacific Gas & Electric Company is hereby required to submit a written statement or explanation to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, with a copy to the Regional Administrator, Region V, and a copy to the NRC Resident Inspector, Diablo Canyon Power Plant, within 30 days of the date
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3-This reply of the letter transmitting this Notice of Violation (Notice).
should be clearly marked as a " Reply to a Notice of Violation" and should include for each violation: (1) the reason for the violation, or, if contested, the basis for disputing the violation, (2) the corrective steps that have been taken and the results schieved, (3) the corrective steps that will be taken to avoid further violations, and (4) the date when fullIf an' adeq compliance will be achieved. time specified in this Notice, an order or a Deman issued to show cause why the Itcense should not be modified, suspended, or Where revoked, or why such other action as may be proper should not be take good cause is shown, consideration will be given to extending the response tim Dated at Walnut Creek, California, this IP day of May 1993
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U. S. NUCLEAR REGULATORY COMMISSION
REGION V
EA Number: 93-107 Report: 50-275/93-11 and 50-323/93-11 Licenses: DPR-80 and DPR-82 Licensee: Pacific Gas and Electric Company (PG&E)
77 Beale Street San Francisco, California 94106 Facility: Diablo Canyon Power Plant (DCPP), Units 1 and 2 Inspection Location: San Luis Obispo Country, California Onsite Inspection: April 5 - 9,1993 In-Office Inspection: April 12 - 28,1993 Inspected by: /I Brepr, M [w Radiatio Specialist
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/ /3 - Cfentz,Senir diation Specialist Date Signed Approved by: / n , ,u 5-/3-93 ClamefjH. Reese, Chief Date Signed FaciTities Radiological Protection Branch Summarv:
Areas Insoected: Routine, unannounced inspection of a followup item (rega. ding the post-accidc apling system) and' occupational exposure controls. Inspection Procedures 83750, 83729, 84750, and 92701 were used.
Results: The licensee's programs for controlling occupational exposure, in the aspects reviewed, were adequate in meeting the licensee's safety objectives. One violation was identified, regarding four instances of failure to implement required controls for entry into posted high radiation areas.
In addition, three apparent violations were noted relating to aspects of the licensee's post-accident sampling system: (1) the licensee failed to implement and maintain a program (per Technical Specification 6.8.4) that ensured the capability to obtain and analyze reactor coolant samples for dissolved hydrogen under accident conditions; (2) the licensee discontinued procedures, training, and calibration of the SENTRY gas chromatograph for dissolved reactor coolant hydrogen analysis, and failed to perform a written safety analysis (per 10 CFR 50.59) of this change; and (3) the licensee failed to implement and maintain a program (per Technical Specification 6.8.4) that er.sured the capability to obtain and analyze samples of radiciodines and particulates in plant gaseous effluents under accident condition _ _ _ _ - - _ _ - - _ _ _ _ - _ - _ _ - _ _ - _ _ _ _ _ _ _ _ _ - .
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a DETAILS 1. Persons Contacted Licensee
- P. Baxter, Chemistry Instructor
- tJ. Boots, Director, Chemistry
- tE. Carlsen, Engineer, Regulatory Compliance
- S. Ehrhardt, Engineer, Radiation Protection (RP)
.i. Fong, Engineer, RP tJ. Gardner, Senior Engineer, Chemistry
T. Grebel, Supervisor, Regulatory Compliance C. Helman, ALARA Engineer
- R. Hess, Assistant Onsite Project Engineer
- T. Irving, General Foreman, RP
- tD. Miklush, Manager, Operations. Services
- J. Molden, Director, I&C
- T. Moulin, Assistant to Vice President
- M. Mosher, Engineer, Quality Assurance (QA)
- W. Rising, Auditor, QA
- R. Snyder, Senior Chemistry Instructor M. Somerville, Senior Engineer, RP tA. Taylor, Senior Engineer
- tR. Thierry, Senior Engineer, Regulatory Compliance
- tE. Wessel, Engineer, Chemistry NRC ,
F. Gee, Resident Inspector M. Miller,' Senior Resident Inspector
+(*) Denotes those< individuals who attended the exit meeting?n; April 9, 1993. The inspector met and held discussions with additional members o the licensee's staff during the onsite inspectio ;
(t) Denotes those individuals who participated in the .teleconference calls on April 29 and 30,199 . Occupational Exposure (83750. 83729)
At the time of the inspection, Unit 2 was in a refueling outage (2R5). l The inspectors examined this program area by performing tours and- '
independent surveys of the facility, observing. work in progress, reviewing procedures and records, and interviewing cognizant personne Observations were made regarding radiological posting and labeling, surveys and monitoring, and control of posted high radiation areas.
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's 2 Radiolooical Postina and labelina While conducting tours of the Auxiliary Building, Containment Building, and various other areas within the licensee's Protected Area (PA), the inspectors observed and verified the licensee's radiological posting and labeling. For those areas observed, radioactive material labels, as well as postings for radiation, high radiation, and radioactive materials areas, were visible, accurate, and curren Surveys and Monitorina The inspectors examined the licensee's surveys, survey logs, and instrument issue logs. In addition, one inspector accompanied a contract RP technician (RPT) on a routine survey of the Unit 1 PA, i Warehouse A, and the I&C instrument calibration facilit The folinwing licensee procedures were reviewed in assessing the license;'s program implementation:
- RCS-7, " Radiation Control Standard - Surveys," Revision 6
- RCP D-500, " Radiation and Contamination Surveys," Revision 9
- RCP D-501, " Issue and Return of Radiation Protection Equipment," Revision 1
- RCP D-510, " Radiation and Contamination Survey Program,"
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Revision 5 The inspectors made specific observations related to survey practices, worker usage of portal monitors and friskers, records of surveys and supervisory review of survey information, inst-ient a avTilability, and adhe1"ence to procedural muirement .=
(1) Survey Practices Upon review of survey logs and instrument issue logs, the inspectors noted that surveys were being performed and documented thoroughly. Routine surveys were being performed at the required frequency. Observation of surveys and discussions with workers revealed that RP personnel had a thorough understanding of the survey program, instrument requirements and limitations, techniques for taking and counting smears, and ALARA practices. The licensee's routine survey practices were found to be adequate for monitoring and posting radiation and high radiation areas and were consistent with 10 CFR 20.201,
" Surveys."
While the inspector was accompanying an RPT on a routine survey of the Unit 1 PA, a discrete radioactive particle (" hot l
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. particle") was found on the asphalt between the north gate of the 115' Radiological Controlled Area (RCA) Backyard and Warehouse Readings with a count-rate meter were in excess of 50,000 corrected counts per minute at 1/2 inch from the particle. The licensee successfully removed the particle from the asphalt, performed a spectral analysis, and disposed of the particle appropriately. Additional RPTs were sent to perform a more thorough survey of the area. No other particles were foun (2) Worker Usaae of Portal Monitors and Friskers I
The inspectors observed worker usage of' portal monitors and friskers at the 85' Auxiliary Building and the 140' Containment Building access control points. The inspectors noted that workers' contamination monitoring practices were adequat '3) Records of Surveys and Supervisory Review of Survey Information The inspectors found the licensee's maintenance of survey records to be acceptable. Licensee supervisory review of survey results was performed efficiently, and survey data and information was disseminated in a timely manner for use in work planning and radiation dose contro (4) Instrument Availability The availability of RP instrumentation was observed at the 85'
Auxiliary Building instrument room. The inspectors noted that few E-140 friskers and telescoping survey instruments were available for use. The RP Foreman on shift indicated that the majority of instruments were in the Containment Building for
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the outage. The RP Eng ;er responsible for coordinating instrument calibration and maintenance with I&C stated that the timeliness of instrument calibration and maintenance still needed improvement, but had improved over previous outage All instruments observed by the inspectors had current calibration and performance test sticker The inspectors noted that Procedure RCP D-500, " Radiation and Contamination Surveys," listed an E-520 as an available instrument for use, but cross-checks against the instrument inventory revealed that the licensee did not possess any E-520 instruments. An equivalent instrument, an ASP-1 audible response with a Geiger-Mueller probe, was available in the licensee's instrument inventor (5) Adherence to Procedural Reouirements The inspectors reviewed survey practices for procedural adherence. Several items were noted:
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(a) Review of survey logs for the period of January 1993 through March 1993 indicated that routine surveys were performed at the frequencies required by Procedure RCP D-510. " Radiation and Contamination Survey Program."
(b) Surveys were thoroughly documented as required by Procedure RCP D-510, " Radiation and Contamination Survey Program."
(c) Other survey practices observed (e.g. smear techniques, instrument use, preparation for surveys) were performed as requi ed by Procedure RCP D-500, " Radiation and Contamination Surveys." Control of Posted Hiah Radiation Areas Technical Specification (TS) 6.12.1 requires that high radiation areas in which the intensity of accessible whole body radiation is greater than 100 millirem per hour (mrem /hr) but i less than or equal to 1000 millirem per hour (mrem /hr) at 18 inches shall be barricaded and conspicuously posted as a high radiation area. Any individual or individuals permitted to l enter such an area must have either: (1) a radiation dose rate i survey meter, or (2) an alarming dosimeter and prior knowledge l of the area dose rates, or :3) accompaniment by an individual l qualified in RP procedures who carries a survey meter, provides l positive control over activities in the area, and performs periodic radiation survey _
The inspectors reviewed four instances in which workers had entered posted high radiation areas (HRAs) without meeting TS
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requirements for HRA controls. The instances, and the licensee's corrective actions for each case, were as follows:
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(1) On October 2, 1992, during the Unit 1 1R5 outage, a contract worker entered the refueling bridge posted HRA to look at a reactor vessel in-service inspection tool. The worker did not have either a survey meter, an alarming dose rate meter, or coverage by an RPT. In addition, the worker had not been briefed on radiation levels present in the HR The worker was observed by a deconner, who called an RPT, who in turn instructed the worker to exit the HR As corrective action for this problem, the worker's RCA access authorization was suspended, and the worker was subsequently terminated. In addition, worker training on this incident was given prior to the Unit 2 2R5 outag (2) On October 26, 1992, during the Unit 1 1R5 outage, two contractor carpenters entered a posted HRA on the 140-foot elevation of the Unit 1 Containment. A nearby RPT
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5 observed these entries, noted the workers' lack of monitoring devices, and promptly removed the workers from the HR As corrective action for this problem, both workers' RCA access authorizations were temporarily placed on hold, and l the event was discussed with their supervision.
l Subsequent training on the event (and on TS requirements for HRA entries) was given to the entire crew on October-27, 199 (3) On March 15, 1993, during the Unit 2 2R5 outage, three contractors were observed by a quality control inspector to be working without proper radiation. monitoring in the posted HRA adjacent to reactor coolant pump (RCP) 2- The workers subsequently stated that they had entered the area by climbing through a narrow gap in the floor grating, and had not known that they were entering the-HRA.
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In investigating this problem, the licensee noted that the workers had received a pre-job briefing on radiation hazards by the RPTs at the 115-foot elevation of Containment; however, the workers had not informed the RPTs of the full scope of the job, and therefore had not been briefed on the radiation levels present in the HRA beneath the floor grating. As corrective action, the workers were counselle ~~
(4) On March 22, 1993, during the Unit 2 2R5 vutage, two contractor workers were found without proper monitoring inside the pressurizer shed posted HRA on the 160-foot elevation of the Unit 2 Containment. The workers had
checked in with the access control RPT, who had forgotten
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to issue them alarming dosimeter As corrective action, both the access control RPT and the workers were counselle During discussions with the inspectors, the RP Director (RPM)
noted that the percentage of error evidenced by these four instances was relatively small in proportion to the number of properly monitored HRA entries during outage periods. The RPM noted, in addition, that in each case subsequent surveys and evaluations had concluded that the workers had not actually entered those portions of the respective HRAs in which the dose field exceeded 100 millirem per hou The inspectors acknowledged these statements, but observed 1'
that, by entering the posted HRA, the workers had in-each case circumvented the final HRA control (i.e., the posted sign-and barricade) and gained access to the high dose fields. The
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inspectors also observed that the licensee's corrective actions thus far had not been effective in preventing recurrence of this problem. The inspectors. concluded _that the failure.to adhere to HRA entry controls, in the four instances noted above, constituted a violation of TS_ 6'.12.1 (50-275/93-11-01).
In response to the inspectors' observations, the RPM' stated: .
that, in an effort to prevent additional improper' HRA entries, i additional controls were_ being evaluated. These controls under-evaluation included: (1) increased emphasis on HRA controls during general employee training; (2) posting additional, even-more conspicuous HRA signs; (3) using different HRA barricades;-
(4) " tightening" the HRA boundaries (i.e., restricting the HRA boundaries to the immediate area of-the high dose fields); and (5) inclusion of more conspicuous markings on radiation work -
permits to remind workers of HRA monitoring' requirements prior:
to beginning a jo With t'e exceptions noted,-the licensee's programs for controlling ,
occupational exposure appeared effective in the areas observed. One violation.of NRC requirements was identifie . Post-Accident Samolina (84750. 92701)
This issue was previously discussed in _NRC Inspection Report 50-275/93-04--
and 50-323/93-04, and was identified as Unresolved Item 50-323/93-04-0 On the basis of the following discussion, the unresolved item is considered close The iiispectors' review of this area was to determine whether the _
l licensee's program for obtaining post-accident samples of reactor coolant and plant gaseous effluents met the requirements of TS 6.8.4.e and the commitments to NUREG-0737 as given in the_ Updated Final Safety Analysis
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Report, Chapter 9.3.2.2, and historical correspondence between PaciD c Gas and Electric Company (PG&E) and the NR In conducting this review, the inspectors reviewed applicable licensee procedures, calculations,
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maintenance records, and relevant historical correspondence, and discussed the post-accident sampling system (PASS) history and performance with various members of the Chemistry Department and licensee managemen Discrepancies were noted regarding two aspects of the licensee's PASS program: (1) sampling and analysis of dissolved hydrogen in reactor coolant; and (2) sampling and analysis of radioiodines and particulates in plant gaseous effluent Samplina and Analysis of Dissolved Hydrocen in Reactor Coolant (1) Recuirements. Criteria. and Commitments TS 6.8.4 requires the licensee, in part, to establish, implement, and maintain a program which will ensure the .j
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capability to obtain and analyze reactor coolant under accident conditions. The program must include (1) training of personnel, (2) procedures for sampling and analysis, and (3)
provisions for maintenance of sampling and analysis equipmen UFSAR Section 9.3.2.2, " Post-LOCA Sampling System," further describes the licensee's methods of implementing this program, including the system capability for quantifying dissolved hydrogen in reactor coolant. The UFSAR commitment states that all sampling and analyses, "as required by NUREG-0737 (November 1980)," can be done within a 3-hour perio NUREG-0737 (November 1980), " Clarification of TMI Action Plan Requirements,"Section II.B.3, provides criteria for postaccident sampling capability. These criteria include:
(a) The capability of providing, within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of deciding to take a sample, quantification of either hydrogen gasas or total dissolved gases in the reactor coolant; (b) The capability of performing backup grab samples for any parameter using inline monitoring; and (c) Use of the dose limits of General Design Criteria (GDC) 19 of 10 CFR 50, Appendix A (i.e., 5 rem whole body, 75 rem extremities) as the design basis for any individual performing postaccident sampling and analysi PG&E Letter DCL-85-047, dated February 1, 1985, committed to having the Diablo Canyon Unit 2 SENTRY PASS operable prior to criticalit Enclosure 1 to this letter ' stated that the SENTRY PASS was "the permanent system for providing post accident
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sampling capability," and reiterated PG&E's commitment to
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meeting the criteria of NUREG-0737,Section II. Finally,10 CFR 50.59 requires, in part, that the licensee shall maintain records of changes to the facility or procedures described in the UFSAR, including a written safety evaluation that provides the basis for determining that the change does not involve an unreviewed safety questio (2) Backaround The SENTRY PASS method of quantifying dissolved hydrogen in reactor coolant consisted of a shielded off-line syste A l small, pressurized sample of reactor coolant liquid was isolated within the system and evacuated to a gas bom The l offgas was routed through a shielded gas chromatograph (GC) to obtain a hydrogen concentration readou After several years of using the SENTRY PASS, the licensee developed a concern regarding moisture carryover from the l
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8 1 liquid sample through the gas bomb and:into the SENTRY'GC; i
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With routine use of the system (for training and operationa checks), the: carried-over moisture periodically wetted down the GC columns, rendering the GC inoperable until the columns could be replaced. Since the SENTRY GC was also used for analyzing
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dissolved oxygen in the reactor coolant and hydrogen in the containment atmosphere, this matter was of additional concer (3) ' System Chanaes As an improvement to.the system, the licensee' decided to ad in-line reactor coolant hydrogen monitors. These monitors,-
called exosensors,-were designed with a gas-permeable membran Hydrogen passing through the membrane reacted with a sulfuric !
acid solution, and this reaction was measured to quantify:the hydrogen concentratio j The -licensee's evaluation of adding the exosensors,- performed in accordance with 10 CFR 50.59; was completed in December' ;
1987. The evaluation stated that-the SENTRY GC would continue to be maintained as- an alternate method of analysis. The evaluation concluded that the UFSAR analysis was. unaffected by ,
this change (i.e., by-. adding the exosensors)'.
The Unit 2 exosensor was installed and plant-accepted on-December 8, 1989, and the. Unit 1 exosensor was installed and plant-accepted on January 15,1990. " After more than a year of .
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observing exosensor. performance in both' units, operation of the !
exosensors was added to the PASS-emergency procedures on March- t 27, 199 ;
Due to continued problems with moisture carryover in the SENTRY GC, the licensee discontinued use of the'GC for' dissolved ;
reactor coola... hydrogen. analysis.(for both Units 1 and 2). On ;
August 18, 1992, this method of analysis was removed from the- 'l PASS emergency. procedures, and training'in this method wa !
discontinued for PASS. operators. Comparisons of SENTRY GC l reactor coolant hydrogen to normal reactor coolant hydrogen . l samples, formerly- used to determine calibration of the' SENTRY GC method, were no longer performed -after this dat (4) NRC Evaluation As previously noted in NRC Inspection. Report 50-275/93-04 and 50-323/93-04, the Unit 2 exosensor performed much more poorly than the Unit 1 exosensor. From August-1,.1992,.to the time o the onsite inspection,.the licensee had considered the Unit 2 exosensor to be inoperable during the following periods:
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9 Total Days Period August'24, 1992 - October 15, 1992- 53 .
November 6, 1992, - December 10, 1992 35-December 31, 1992 - February 12, 1993 44 February 25, 1993 - April 9, 1993 44 l
During the same period, the Unit 1 exosensor was only ]
inoperable for brief periods of routine maintenanc The inspector noted the following deficiencies related to the ,
licensee's PASS capabilities for reactor coolant dissolved
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hydrogen:
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(a) By invalidating the procedures, training,_ and calibration for the SENTRY GC method of analysis, this' method no longer met the program requirements.of TS 6.8.4. Despite this fact, the SENTRY GC method was still listed as an alternate method of analysis 'under Equipment Control l '
Guideline (ECG) 11.1, " Post Accident Sampling ~ System."
(b) Other alternate methods given in ECG 11.1 included normal laboratory analysis of liquid grab samples obtained. from the PASS sample lines or normal reactor coolant sample-lines. Using these methods, however, the licensee was unable to perform sampling. and analysis .within the design !
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basis dose criteria of GDC 19 under accident conditions (as outlined by NUREG-0737 criteria).-
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(c) As a result, for both Units 1 and-2, inline monitoring of - .
reactor coolant dissolved hydrogen was not' supplemented by backup grab sampling capability, as outlined by NUREG-0737 - l criteri l en .. 3 J l ,
(d) During the periods of exosensor inoperability in Unit 2,- '
no method of sampling and analyzing reactor coolant ~
dissolved hydrogen under accident conditions met the'
criteria of NUREG-073 ,
(e) The licensee's decision to discontinue the-procedures, training, and calibration of the SENTRY GC for reactor coolant dissolved hydrogen analysis was not within-the scope of the 10 CFR 50.59 evaluation performed in December 1 1987. The licensee had not performed any additional evaluation of this change to the PASS capabilit (5) Licensee Evaluation and Corrective Action The inspectors noted that enclosures-to the licensee's February-1, 1985, commitment letter had listed liquid grab samples (drawn from either the PASS lines or normal reactor coolant-sample lines) as alternate methods of meeting NUREG-0737 PASS l \
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I criteri In discussions with the inspectors, the licensee l stated that PG&E had never performed evaluations to' demonstrate l l that these alternate methods could meet the design basis dose
! criteria'of GDC-19. The licensee had incorrectly assumed that normal laboratory analysis of liquid grab samples were always an available alternative to the exosensor or SENTRY G l During an April 14, 1993, conference call, the licensee discussed proposed corrective actions with members of NRC Region V management. On April 15, the licensee submitted PG&E Letter DCL-93-089, which comitted to the following corrective actions:
(a) The SENTRY GC method of quantifying reactor coolant dissolved hydrogen would be upgraded by adding a liquid l coalescing filter to reduce moisture carryover and extend thermal conductivity detector lif This upgrade would be .
completed in Units 1 and 2 by April 21, 1993. The letter i noted that the SENTRY GC method was considered a remote grab sample.
l l (b) The licensee had already revised applicable procedures to l include the SENTRY GC method of quantifying reactor l coolant dissolved hydrogen. In addition, calibration and
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interim training for this method would be completed by April 21, 1993. Training of.all PASS assigned shift chemistry and RP technicians would be completed by May 15, :
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(c) The Unit 2 exosensor wouu be'returnea to operable status l prior to restart of Unit !
During an April 21, 1993, conference call, the liter *ae stated i that the design-change installing coclescing filters in the SENTRY GC lines had been completed in both units. 'In addition, the licensee stated that interim training and calibration had been completed, and that the Unit 2 exosensor had been repaired and declared operable. Finally, the licensee stated that calibration of the exosensor would be verified against normal laboratory analysis of reactor coolant upon restart of Unit 2.
(6) Conclusion
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The inspectors determined that the licensee had failed to
! implement and maintain a program that ensured the capabilities for sampling and analysis required by TS 6.8.4, in the
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following aspects:
(a) At the time of discontinuing the procedures, training, and calibration for the SENTRY GC method of quantifying reactor coolant dissolved hydrogen, the licensee had l
removed grab sampling capability as a backup to inline
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j monitorin (b) During the periods of Unit 2 exosensor inoperability,.no alternate method was available capable of quantifying ,
reactor coolant dissolved hydrogen under the specified accident conditions without exceeding GDC-19 dose l criteria.
' The inspectors concluded that these failures. constituted an apparent violation of TS 6.8.4 (50-323/93-11-02). The inspectors concluded, further, that the failure to' perform a-written safety evaluation of the effects of invalidating the SENTRY GC method of quantifying reactor coolant dissolved hydrogen constituted an apparent violation of 10 CFR 50.59.(50-523/93-11-03).
In an effort to establish the overall safety consequence.of these deficiencies, the inspectors noted that normal laboratory analysis of reactor coolant dissolved hydrogen while meeting GDC-19 dose criteria was possible under some accident conditions, and that the conditions outlined in NUREG-0737 represented worst-case accident' conditions. The inspector also noted, however, that the licensee's PASS equipment control guidelines and surveillance test procedures did not establish the severity of the accident for which these-" alternate-methods" could be use '
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In addition, the inspectors reviewed portions of the licensee's.
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emergency plan and discussed the use of the reactor coolant dissolved hydrogen analysis with applicable members of.the licensee's emergency organization. .The inspectors noted.that :
this analysis was not used as a. primary'means of assessing core damage under accident ~ nditions.. Rather, it was considered
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supplemental information used in verificati.on and long-term analysis of core condition Samplina and Analysis of Radioiodines and Particulates in Plant Gaseous Effluents 1 (1) Reouirements. Criteria. and Commitments TS 6.8.4 requires the licensee, in part, to establish,
-l implement, and maintain a program which will ensure ,the .
capability to obtain and analyze samples of radioiodines and -l I
particulates in plant gaseous effluents under accident conditions. The program must include (1) training of -
personnel, (2) ~ procedures for sampling and' analysis,o and (3)
provisions for maintenance of sampling and analysis equipmen ]
UFSAR Section 11.4 further describes the licensee's methods of implementing this program, including a basic description of the-use and functions of the midrange plant vent. iodine monitor j l
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l (RE-32) and high-range plant vent iodine sampler (RX-40).
l In a letter to the NRC dated April 15, 1982, PG&E committed to l meeting the criteria of NUREG-0737,Section II.F.1, Attachment 2 for the gaseous iodine and particulate sampling system in Diablo Canyon Unit 1. In a letter to the NRC dated July 21, 1982, PG&E reiterated this commitment, and stated that RX-40 l had been made operational as of July 8,198 NUREG-0737 (November 1980), " Clarification of TMI Action Plan I
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Requirements,"Section II.F.1, Attachment 2, provides criteria for the licensee's post-accident capability of quantifying radiciodines and particulates in plant effluents. These i criteria include: 1 (a) Concentration of 100 uCi/cc of gaseous radioiodines and particulates, deposited on the sampling media; l
(b) 30 minutes of sampling time; ;
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(c) An average gamma energy of 0.5 MeV; and (d) Design of the sampling system such that plant personnel could remove samples, replace sampling media, and ,
, transport the samples to the onsite analysis facility with '
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radiation exposures not in excess of the GDC-19 dose criteri In addition,Section II.B.2 of NUREG-0737 provides the source term to be used in reviewing plant shielding design and in determining the accessibility of vital. areas during post-accident operation (2) Backarou:J As originally installed, the licensee had several monitors and/or samplers capable of quantifying plant vent radiciodines '
and particulates under various conditions: )
l (a) RE-24, the normal range monitor, was located on the 115' l elevation of the Fuel Handling Building in the Plant Vent ;
Roo RE-24 provided a local readout and alarm function, as well as a . emote alarm in the Control Room. In )
addition, RE-24 was equipped with a particulate filter and i silver zeolite iodine cartridge that could be removed and analyzed as a grab sample. The approximated useful range of RE-24 was from 1 E-7 microcuries per cubic centimeter (uti/cc) to 1 E-4 uCi/c (b) RE-32, the midrange monitor, was located adjacent to RE-24 on the 115' elevation of the Fuel Handling Building. RE-32 provided remote monitoring, strip chart recorder, and I
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. alarm functiocs in the Control Room, and could be remotely operated and purged. In addition, RE-32 was equipped with a particulate filter and silver zeolite iodine cartridge that could be removed and analyzed as a grab sample. The approximated useful range of RE-32 was from 1.3 E-7 uCi/cc to 3 E-3 uCi/c (c) RX-40, the high-range sampler, was located outside at the 85' elevation against the north wall of the Fuel Handling Building. RX-40 was provided with sampling functions only (no monitoring or alarm functions), and was operated remotely from the Control Room. A particulate filter and silver zeolite iodine cartridge was housed in a shielded lead pig assembly, which could be extracted into a larger, lead-shielded cart for transport to the Technical Support Center for analysis as a grab sample under accident conditions. The approximated useful range of RX-40 was from 1 E-9 uCi/cc to 1 E+2 uCi/c '
(3) System Modifications The licensee was in the process of installing upgrades in both units for various process and effluent radiation monitors. As part of this upgrade program in Unit 1, RE-24 had been removed, and a portable RADEC0 pump assembly had been installed in the former RE-24 sampler location. This assembly was provided with a particulate filter and silver zeolite iodine cartridge for periodic grab samplin _
In review of Unit 1 Control Room logs, the inspectors noted that the licensee had taken both RE-32 and RX-40 out of service for a 12-day period from February 26 - March 9, 1993. As part of continued monitor upgrade work, the monitors had been in variousaconditions of inoperability during this +ime (e.g., ,
clearances hung with supply breakers open, leads lifted, heat trace inoperable). ,
(4) NRC Evaluation The inspectors noted that log entries on February 26, 27, and 28 gave a list of monitors out of service and then stated, "No alternate sampling available for PASS" [ referring to the plant vent radiciodine/ particulate monitoring capability only].
l Although subsequent log entries recorded the same monitors as l
being out of service from February 26 - March 9, no entries l
regarding PASS sampling capability were made after the February 28 entry.
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The inspectors asked the chemistry engineer responsible for i PASS what alternate means had been provided for post-accident sampling of plant vent radiciodines and particulates during the period in question. The chemistry engineer stated that, based l
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(STP) G-14, " Operability Determination of Post Accident ,
Sampling Program," listed RE-24 as an alternate post-accident sampling point for plant vent radioiodines and particulate RE-24 was also listed as an alternate sampling point in Equipment Control Guideline (ECG) 11.1', " Post Accident Sampling System." The inspectors noted, however, the following concerns:
i (a) The grab sample cartridge at the RE-24 location had not i been provided with any shielding. In addition, the RADECO l
pump temporarily attached to RE-24 sampling lines required i local operation. Under post-accident conditions, this i
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would require prolonged exposure in a location approximately 2 - 3 feet from the plant ven (b) Accessing the RE-24 sampling location on the 115' Fuel i Handling Building would require passing the fil % banks on the 115' hallway. Under post-accident conditions, i l these filter banks would become highly radioactive, ;
creating a considerable radiological source term in the !
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hallway.
l (c) Time and motion studies had been performed to ensure the l capability to obtain and analyze a sample from RX-40 under l accident conditions; however, no such studies % d been performedafor sampling from RE-2 .#
(d) Despite listing RE-24 as an acceptable alternate for RX-40 in the above procedures, the licensee had never determined that the RE-24 design allowed collecting a sample under accident conditions within the GDC-19 dose criteria, nor was such a determination made during the February 26 -
March 9 period to justify dependence on RE-2 At the exit interview on April 9, 1993, the inspectors asked l the licensee to perform a dose calculation to determine whether
! samples could have justifiably been obtained under specified l
accident conditions using the temporary RADEC0 pump at the RE-l 24 locatio (5) Licensee Evaluation
! On April 13, 1993, the licensee provided the inspectors with a calculation of source term and resultant dose entitled
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" Extemporaneous Use of RE-24's 'Old' Sample Line for Post Accident Sampling in Connection with CAP E-2 & EP RB-12 in the Event RX-40 Is Not Operable." This study accounted for exposure from (1) the sampling cartridge, (2) the filter. bank hallway, (3) outside area exposure, and (4) plant vent exposure while obtaining the sample. The study concluded that an individual obtaining and analyzing such a sample would receive approximately 0.48 rem whole body dose and 4.8 rem extremity dose, without the use of shielded sampling or transport cask Based on the above study, the licensee concluded that the use of this proposed sampling location would have provided an acceptable means of obtaining and analyzing radioiodines and particulates in plant gaseous effluents under accident condition (6) Further NRC Evaluation and Conclusion The inspectors reviewed the licensee's study, and notea the following concerns:
(a) The licensee's study had not used the applicable source term provided in NUREG-0737,Section II.F.1, Attachment 2 (as committed in the April 15, 1982,- letter), for evaluating the dose from.the radioiodine cartridge and particulate filte Using the applicable source term from NUREG-0737, the
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inspectors recalculated the whole body dose from handling the sampling cartridge for 90 seconds to be approximately 25 re (b) The licensee's r+"iy had not used the enalysis methods of
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NUREG-0737,Section II.B.2 for evaluating accessibility of the Plant Vent Room on the 115' elevation of the Fuel Handling Building during post-accident operation (c) Emergency Procedure (EP) RB-12, "Mid and High Range Plant Ver.t Radiation Monitors," Appendix 1, Section 4.b, stated the following:
Should radioactivity levels of plant vent effluents rise to the operating level of the RE-29 monitor, it would be virtually impossible to detect radioiodines in the presence of the noble gases. Because of the high radiation levels from the plant vent itself, personal exposures from entering the monitor area to obtain grab samples would be prohibitive. For this reason, a separate iodine grab sampler (RX-40) has been installe '
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(d) The inspectors concluded that the licensee's analysis was inadequate, and that the licensee's conclusion regarding the adequacy of this sampling location was in error. In subsequent discussions held on April 29, 1993, the licensee acknowledged the inspectors' observations. The i licensee stated that an additional possible sampling i
location would have been from the new RE-24 and RE-24R l monitors, which had not yet been declared' fully operational at the time in questio The inspectors concluded that, from February 26 to March 9, 1993, the licensee had failed to implement and maintain a program which ensured the capability to obtain and analyze samples of radioiodines and particulates in the plant gaseous effluents under accident conditions, constituting an apparent violation of TS 6.8.4 (50-275/93-11-04).
Regarding the overall safety consequence of these deficiencies,
! the inspectors noted that the Plant Vent Room on the 115'
l elevation of the Fuel Handling Building would be accessible l
while meeting GDC-19 dose criteria under some accident conditions, and that the conditions provided in NUREG-0737 represented worst-case accident conditions. The inspectors also noted, however, that the licensee's PASS equipment control l guidelines and surveillance test procedures did not establish l the severity of the accident for which this " alternate" sampling location could be use In addition, the inspectors noted that Emergency Procedure (EP)
RB-9, " Calculation of Release Rate;" proposed using the'RX-40 sampling data as one input to EP RB-ll, " Emergency Off-Site Dose Calculations." The inspectors noted, further, that these calculations of release rate and resultant off-site dose could be used in .a.ablishing protective action recommendations under certain accident condition Additional Observations Related to PASS The inspectors noted that the licensee had incorrectly assumed that
" alternate methods" of post-accident sampling, when relied on, did not need to meet GDC-19 dose criteria. As a result, the licensee had not, as a rule, conducted time-and-motion studies for the alternate methods of any PASS analysis listed in the equipment control guideline or surveillance test procedure. The inspectors noted that additional deficiencies could result from dependence on these unanalyzed alternate method Summary The inspectors concluded that the licensce's PASS program required additional management attention regarding the criteria to be met when changes were made to the method for obtaining and analyzing
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samples. Three apparent violations .were identifie . Exit Interview The inspectors met with' members of licensee manag'ementi at the_ conclusion of the onsite portion of-the inspection on April 9,.1993. ~0n April 29-
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and 30, at the-conclusion of the in-office portion _ of the inspection,--
additional ~ discussions were held with~the individuals noted in Section .
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