IR 05000344/1986009

Insp Rept 50-344/86-09 on 860317-21.Violation Noted:Failure to Rept Cause of Inoperable Radioactive Effluent Monitoring Instrument Channel in Semiannual Radiological Environ Rept

ML20141H878
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	04/11/1986
From:	Hooker C, Yuhas G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML20141H870	List: IR 05000344/1986009 ML20141H864 ML20141H872
References
50-344-86-09, 50-344-86-9, GL-85-08, GL-85-8, IEIN-85-081, IEIN-85-087, IEIN-85-092, IEIN-85-81, IEIN-85-87, IEIN-85-92, IEINB-85-81, NUDOCS 8604250226
Download: ML20141H878 (10)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION V

Report N /86-09 Docket N License N NPF-1 Licensee: Portland General Electric Company 121 S. W. Salmon Street Portland, Oregon 97204 Facility Name: Trojan Nuclear Plant Inspection at: Rainier, Oregon

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Inspection conducted: March 17-21, 1986 Inspector: N./ a #[//

C. A. Hooker, Radiation Specialist Date Signed Approved by: kkp 'Olt/V6 G. P. Tuhas, Chief Dat'e Signed FaciliSie)s Radiological Protection Section Summary:

Inspection on March 17-21, 1986 (Report No. 50-344/86-09)

Areas Inspected: Routine, unannounced inspection of licensee action on previous inspection findings; solid waste; preparation for refueling outage; facility tour; steam generator tube leakage; and followup on IE Information Notice Inspection Procedure 83729, 84722, 84724, and 92703 were covere Results: Of the areas inspected, one violation was identified in one area:

TS 3.3.3-11 ACTION (b), failure to report the cause of an inoperable radioactive effluent monitoring instrument channel in the Semiannual Radiological Environmental Report (paragraph 5 ).

8604250226 860411 PDR ADOCK 05000344 G PDR

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Details Persons Contacted PGE Personnel l *R. P. Schmitt, Manager, Operations and Maintenance

• J. D. Reid, Manager, Plant Services

• C. H. Brown, Manager, Quality Assurance (QA) Operations

• T. O. Meek, Radiation Protection (RP) Supervisor

• G. I.. Rich, Chemistry Supervisor

• S. A. Bauer, Onsite Regulation Engineer L. D. Larson, Radwaste Supervisor R. A. Reinart, Instrument and Control (I6C) Supervisor E. A. Curtis, QA Engineer NRC Resident Inspectors

• S. A. Richards, Senior Resident Inspector G. Kellund, Resident Inspector

• Denotes those present at the exit interview on March 21, 198 In addition to the individuals-identified above, the inspector met and

held discussions with other members of the licensee's staf . Licensee Action on Previous Inspection Findings (Closed) Open Item (GL-85-08): Inspection Report 50-344/85-33, paragraph 15 documented that the licensee had made no decision with respect to Generic Letter (CL) No. 85-08, 10 CFR 20.408 Termination Report - Format. During this inspection, the inspector reviewed the licensee's response letter dated February 21, 1986, to the NRC concerning GL 85-0 The licensee is implementing an in-house computer based record system for occupational radiation exposures and has planned to transfer all 10 CFR 20.408 termination data prepared after January 1, 1987, to the NRC by the electronic transmission pilot program with the data file layout referenced in GL 85-08. This matter is considered close . Solid Waste Audits No audits in this area were identified since the last inspection (50-344/85-09).

No violations or deviations were identifie Changes The licensee had nearly completed construction of a new Fuel Building Annex that will be used for sorting and compaction of dry waste. The facility will also be used for sorting and packaging of contaminated protective clothing. This new facility change was

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reviewed for compliance with 10 CFR 50.59, Technical Specifications (T.J.)sndtherecommendationsofNRCRegulatoryGuide (RG) 1.14 The. licensee's existing dry waste compaction room located in the fuel building (93 ft. elevation) is limited on space for processing of dry waste and has minimal ventilation control capabilities. The new annex (45 ft. elevation) will provide a larger area for the compaction of dry waste and improved ventilation control capabilities. The construction of the new dry waste proceasing facility represents an improvement to the solid waste management syste The inspector reviewed the licensee's Request for Design Change (RDC) No.81-049, Rev. 1, dated June 18, 1981 and the Preliminary Design and Safety Evaluation approved by the Plant General Manager on November 13, 1984. The inspector also made a tour of the new facility. From review of these documents and the facility tour the following observations were noted:

Radioactive liquids will not be processed or stored in the new facilit *

The waste processing will consist of a new dry waste drum compactor and a stainless steel air exhaust hood for waste and contaminated laundry sorting. The waste compactor is equipped with a HEPA filtered exhaust system. The exhaust air from the annex will be connected to the existing Fuel / Auxiliary Building ventilation system, thereby precluding a new effluent release pathwa *

Fire protection will ' ec provided with an automatic sprinkler system. A catch sump with sufficient capacity to contain one sprinkler discharging for 20 minutes at 50 gallons per minute will be provided with a sump pump for discharging liquid. The sump will be grab sampled and discharged to the liquid radwaste system if radioactive contaminants are identifie If the sump water is free of contaminants, it will be discharged to the environmen *

A constant air monitor and personnel contamination survey instruments will be utilized in the are There is no change in the method for processing of dry radioactive wast i l

Based on the above observations, the licensee's evaluation and facility tour, the inspector noted that the design and construction appeared to be in accordance with applicable NRC RGs 1.143, 8.8 and 8.10. The licensee's evaluation concluded that the facility change could be made in accordance with 10 CFR Part 50.59, since the change did not constitute an unreviewed safety question or T.S. change Applicable sections of the FSAR were addresse No violations or deviations were identifie J

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3 Processing and Storage The licensee continues to use the services of a contract vendor for onsite solidification of radioactive waste. The licensee used offsite vendors for waste stream sample analysis for identification of radionuclides and concentrations within the radioactive waste stream Split samples of the spent resin storage tank and primary coolant are sent to two independent vendors for sample analysis for additional quality control checks. The following licensee procedures were reviewed:

RPMP-2 " Radioactive Waste Dramming" RPMP-2-1 "Radwaste Drumming - Absorbed Liquids" RPMP-2-2 " Determination of Liquid / Absorbed Ratio" RPMP-4 " Determination of Radioactive Material Shipping and Waste Classification" RPMP-5 " Sampling Program to Determine Isotopic

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Concentration and Scaling Factors for Classification of Low-Level Solid Radwaste" No violations or deviations were identifie Disposal of Low-Level Wastes Records of the following solid waste shipments were examined:

January 23, 1986, shipment No. 86-2, 84 drums of condensate powdex resi *

January 30, 1986, shipment No. 86-5, two liners of steam generator blowdown dewatered resin *

February 11, 1986, shipment No. 86-7, 100 drums of compacted j dry wast The inspector verified that the licensee classifies waste pursuant to 10 CFR 61.55; verifies that waste meets the characteristics of 10 CFR 61.56; and prepares a waste manifest and marks packages in accordance with 10 CFP. Part 20.311. The waste receiver noted no discrepancies upon receipt of the licensee's waste shipments and no shipments had been lost or damage No violations or deviations were identifie . Refueling Outage Preparations The inspector discussed licensee preparations and plans for the upcoming refueling outage with the EP supervisor. The licensee had contracted for an additional 55 to 60 RP technicians to augment the present plant staff for routine refueling outage tasks and steac generator tube leak repair I

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The inspector was informed that the contract RP technician training would consist of the licensee's general employee training (GET) and RP procedures. The licensee's steam generator (S/G) mockup will also be used for those involved in S/G examination and tube repair wor The inspector also attended and observed GET being given to various temporary craft and professional personnel that will be performing work at the plant during the refueling outag No violations or deviations were identifie . Primary-to-Secondary Leak Rate Through the Steam Generators (PSLR-S/Gs)

The purpose of this examination was to review the circumstance involving a sudden decrease in the licensee's PSLR-S/Gs on March 12, 1986, and review the methods used for determining this leak rat '

Background Prior to November 1985, the licensee's PSLR-S/Gs was less than 10 gallons per day (gpd). As of the first week in November 1985, the licensee has experienced a steady increase with the current leak rate being about 340 j gpd. On March 12, 1986, the Region V office was informed that the leak rate had suddenly decreased from 300 gpd to 70 gpd. This decrease was based on readings from a new, calibrated, flow rate meter that the licensee had installed for measuring the offgas flow rate in the condenser air ejector (CAEJ) monitoring system. The Region V office also learned that the licensee was using the condenser offgas radioactive effluent release rate as the method for determining the PSLR-S/G CAEJ Flow Rate Measurements Based on TS review, discussions with licensee representatives, review of selected documents and facility tours, the following observations were note *

On December 20, 1984, the Commission issued License Amendment No. 99, effective January 1, 1985, that included numerous changes to the licensee's Radiological Effluent T.S. to comply with Appendix I of 10 CFR Part 5 Technical Specification l

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Limiting Condition for Operation (LCO) 3.3.3.11 (Amendment No. 99) reyttred, in part, specific effluent monitoring instrument ch.nnels shown in TS table 3.3-13 to be operabl Table 3.3-13, item 5.d. listed the effluent system flow rate measuring device (flow transmitter FR-3100) as being a required instrument channel for the CAEJ monitoring system during releases via this pathway. Table 3.3-11 Notatica (b)

did not require FR-3100 to be operable until the moisture removal system was installed and operable for PRM-6 "CAEJ Radioactive Effluent Monitoring System" Action Statement N for FR-3100 being inoperable stated, in part, that effluent releases via this pathway may continue, provided the flow rate is estimated at least once every 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> .

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Through discussions with licensee representatives, the inspector learned that the moisture removal system for PRM-6 became fully operational in July 1985. The inspector was informed that FR-3100 had been inoperable since installation in 1984 due to inadequate design. With FR-3100 being inoperable, the licensee was using an installed narrow range (0-22 cfm)

rotameter to determine the CAEJ offgas flow rate. The inspector noted, by review of the control room shift data log (POT-24-1-0), that readings fron FI-3184 were being recorded each shift to estimate the offgas flow as required by Action Statement No. 27 mcationed abov *

The inspector was informed that upon getting the PRM-6 operational, the licensee became aware that FI-3184 had not been calibrated since installation in 1981. Therefore, it was decided to purchase a new calibrated rotamete *

The inspector noted that on July 3, 1985, the licensee generated purchase order No. NQ-00186 to acquire a new rotameter. In late December, 1985, the licensee received the new rotameter; however, it was not installed at this time since the vendor had not supplied the calibration certificate. On March 11, 1986, the licensee installed the new rotameter for FI-3184 after receipt of the calibration certificat *

With the new rotameter, the licensee observed that the CAEJ offgas flow rate (1.5 to 4.0 cfm) was significantly lower, when compared to the flow rates observed with the old rotameter ( to 7.5 cfm). As a result of the lower indicated FI-3184 flow rate and the licensee using the CAEJ offgas radioactive effluent release rate to determine the PSLR-S/Gs, a significant drop in the leak rate (300 gpd to 70 gpd) was observed by the licensee. Due to this anomaly, on March 13, 1986 the old flow meter was checked and indicated good agreement and linearity with the licensee's calibrated secondary standard. On March 14, 1986, the licensee removed the new rotameter and installed the old one that had been calibrated against the licensce's standard. The indicated flow rates were as before (5.5 to 7.5 cfm).

On March 17, 1986, the licensee checked the new rotameter against their standard. When checked, the new rotameter indicated poor agreement and was non linear. The licensee decided to leave the old meter installed based on the calibration data obtained with their standard. The inspector noted that the licensee's secondary standard was last certified on August 29, 198 *

The I6C supervisor informed the inspector that he had concerns in regard to their rotameter flow rate calibration techniqu The I&C supervisor stated that they calibrated the FI-3184 rotameter by venting it to atmosphere, and do not perform tests to determine inline supply and back pressure or moisture content in the CAEJ offgas flow line. The inspector was also

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informed that this matter has been addressed to the engineering department for resolution. The I&C supervisor stated that based on their calibration of the two meters, he was certain that the old rotameter currently installed was more accurate than the new rotameter purchase Licensee resolution as to the proper calibration procedure FI-3184 will be examined in a subsequent inspection -

(50-344/86-09-01, Open) .

At the exit meeting on March 21, 1986, the inspector expressed concerns that for years the licensee has used an uncalibrated instrument (FI-3184) to measure a key parameter (CAEJ flow rate) being used to meet T.S. requirements. (50-344/86-09-02, Unresolved).

Unresolved Item An unresolved item is a matter about which nere information is required in order to ascertain whether it is an acceptable item, an open item, a deviation, or a violatio Based on further T.S. review, the following additional observation was made:

Technical Specification Limiting Condition for Operation 3.3.3.11, Action (b) requires, in part, that with inoperable instrument channels identified in Table 3.3-13, items 1, 2, 3c, 4 and 5 not returned to OPERABLE status within 30 days, identify the cause of the inoperable channels in the Semiannual Radiological Environmental Report in lieu of any other repor *

As mentioned above, T.S. 3.3.3.11 LCO, Table 3.3-13, item 5.d, Notation (b) required FR-3100 to be operable when the moisture removal system for PRM-6 became operational. It was also mentioned above that PRM-6 became operational in July 1985 and FR-3100 has not been operational since installation in 198 Based on discussions with the Chemistry Supervisor, and representatives of the corporate office on March 20, 1986, the inspector learned that the licensee had not reported that FR-3100 was inoperable in their Annual Report dated February 25, 1986, which included the Semiannual Radiological Effluent reporting period of July 1, 1985.through December 31, 1985. Failure to report that FR-3100 was inoperable was identified as an apparent violation of T.S. 3.3.3.11 l (50-344/86-09-03). I l

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At the exit meeting on March 21, 1986, the licensee provided the inspector a Special Report, dated March 21, 1986, reporting that FR-3100 was inoperable and had not been reported as required by T.S. 3.3.3.11. This Special Report was generated i

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after the inspector brought this matter to the licensee's attentio One apparent violation was identified in this are B. Licensee Procedure to Determine the PSLR-S/Gs Technical Specification LCO 3.4.6.2 establishes the limits for reactor coolant system (RCS) leakage. Item C of T.S. 3.4.6.2 limits the PSLR-S/Gs to 1 gpm through all S/Gs and 500 gpd through any one S/ Technical Specification Surveillance Requirement 4.4.6.2.1 lists five methods that shall be used to determine RCS leakage. However, none of these methods specify a technique to determine the PSLR-S/Gs. Of the five items listed, only item, d. " Performance of a RCS water inventory balance at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during steady state operations," could reasonably be expected to be the method to determine the PSLR-S/G The inspector held discussions with licensee representatives, reviewed procedures; Chemistry Manual Procedure CMP-1

" Primary-to-Secondary Leak Rate," Revision 3, dated September 11, !

1984; and Periodic Operating Test POT-1-3 " Leakage Evaluation," '

Revision 11, dated March 21, 1985, to determine the methods used by the licensee in order to comply with T.S. 3.4.6.2.d requirement Based on these discussions, and procedure review, the inspector learned that:

Procedure POT-1-3 provides the methods and procedures used to evaluate leakage from the RCS in accordance with .4.6.2.1.d. This procedure uses the mass balance to determine the primary average leak rate to include: the RCS; pressurizer; volume control tank; RCS sample line; and the letdown and charging lines. In addition, the RCS leak paths to the pressurizer relief tank and reactor coolant drain tank are monitored to distinguish between identified and unidentified i leakege. No mention of PSLR-S/Gs were noted in the contents of j the procedur The licensee felt that using the T.S. 4.4.6.2.1.d method they could see about 0.1 gpm or 144 gp Based on subsequent licensee review of data comparing S/G tube leak rate, there is little confidence in the ability to detect a 0.1 gpm change in the leak rate. The licensee has not performed an error analysis associated with the T.S. 4.4.6.2.1.d metho *

Procedure CMP-I provides the procedures used in meeting T.S. 3.4.6.2.C (PSLR-S/Gs) requirements and other methods used for compariso It was noted in the opening statement of the procedure that the method of determination of primary-to-secondary leakage is not specified in tha _ . _ _ __ _ _ _ . - _ - . _ . . .._

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In further review of this procedure,Section I " Leak rate Quantification," provided the following information:

1) Detection of PSLR-S/G is accomplished by radiochemical measurements either from radiation monitors and samples at the air ejector, or samples taken from the blowdown or main stea ) For leak rate determination, radioactive tritium and ,

l gaseous isotopes are best for quantifying the leak, because they are volatized in the steam generato Tritium is uniformly distributed throughout the condensate and feed system and' noble gases are removed by the air i ejectors. Decay from these isotopes may be neglected I since tritium has a half life of 12.3 years and noble gases have a short residence time in the secondary system. ,

l 3) Iodine-131 with a 8-day half life builds up in the S/G water and provides a good indication of a primary system lea Iodine in the S/Gs is very dependent on the blowdown rate, so accuracy by this method is dependent by the accuracy in determing the blowdown rat ) Xenon-133 and 135, and argone-141 are generally used for leak rate calculations. Tritium is used for a backup

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method, but due to the delay time of using this method, it is not used. Since xenon-133 is the highest concentration in the reactor coolant and the condenser offgas, the statistical accuracy of its determination is greates Xenon-133 leak rates are used for compliance with T.S. and other methods are used for compariso ) Following power changes, all isotopes experience transient conditions between the reactor coolant and secondary water. This causes inaccurate leak rate determinations until equilibrium concentrations can be reestablishe ) Iodine concentrations will vary depending on fuel defect but will stay in close correlation between primary and secondary water than xenon-133 since the iodine leak rare is more dependent on S/G blowdown than decay. The dependency on blowdown means that the S/G concentration will follow the reactor coolant closer since it is less decay dependent during the early stages of the transien However, iodine leak rates lag the xenon leak rate during sudden changes in the primary-to-secondary leakage, so icdine cannot be used for compliance with *

The inspector was informed by the Chemistry Supervisor that, they could see a PSLR-S/G of about 0.04 gpd. This value is based on the xenon-133 lower limit of detection release rate in

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Based on the above observations, since the T.S. do not specify the method required to determine leakage through any one S/G, the licensee has chosen'a method that appears to be relisble and fairly accurat However, this matter will be discussed with the Office of Nuclear Reactor Regulation . Facility Tour

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The inspector toured various areas of the auxiliary and turbine buildings, and the outside radioactive waste storage area. The inspector made independent measurements using NRC ion chamber S/N 2691 due for calibration April 15, 198 The inspector observed that all radiation areas and high radiation areas were posted as required by 10 CFR Part 20, and access controls were-consistent with T.S. 6.12 and license procedure No violations or deviations were identifie . Followup on Information Notices The inspector verified that the licensee had received, reviewed and was taking or had completed-action on Information Notices 85-81, 85-87 and 85-9 No violation or deviations were identifie . Exit Interview The inspector met with those individuals denoted in report Paragraph 1 on March 21, 1986. The scope and findings of the inspection were summarized. Regarding the apparent violation identified in sub-paragraph ;

5.A, the licensee presented immediate response and corrective actions to i

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prevent future violation l