IR 05000483/1993018

Insp Rept 50-483/93-18 on 931004-08.No Violations Noted. Major Areas Inspected:Radiation Protection & Outage Planning Programs,Chemistry Controls & Effluent Releases

ML20059E085
Person / Time
Site:	Callaway
Issue date:	10/26/1993
From:	Cox C, Michael Kunowski, Nirodh Shah, Snell W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20059E059	List: IR 05000483/1993018 ML20059E055
References
50-483-93-18, NUDOCS 9311030093
Download: ML20059E085 (6)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Report No. 50-483/93018(DRSS)

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Docket No. 50-483 License No. DPR-30 Licensee: Union Electric Company Post Office Box 149 St. Louis, M0 63166 Facility Name: Callaway Nuclear Power Station Inspection At: Callaway Site, Callaway County, Missouri Inspection Conducted: October 4-8, 1993 Inspectors: [ L M. K'unowsfi #

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Approved By: /A Im bN William Snell, Chief so/zc/M Date Radiological Programs Section 2 .

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Inspection Summary Inspection on October 4-8. 1993 (Report No. 50-483/93018(DRSS))

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Areas Inspected: Special, announced inspection of Callaway Nuclear Power -

Station's radiation protection and outage planning programs (Inspection Procedure (IP) 83750), chemistry controls (IP 84750), and effluent releases (IP 84750). The inspection also included tours of the auxiliary and radwaste buildings, containment, and the spent fuel pool area (IP 83750).

Results: The radiation protection program appeared to be effective in implementing the requirements of the regulations. Housekeeping practices were very good for a facility undergoing a refueling outage. Radiation Protection (RP) and station upper management were noted touring the auxiliary building and containment routinely. Excellent coordination was noted between the chemistry and operations group in the controlled shutdown for the refueling outage. Chemistry controls for the shutdown were executed as required by the failed fuel action plan which resulted in overall shutdown radiation levels similar to the previous refueling outage. Improvements were noted in the coordination of the RP group and the outage planning group for incorporating ALARA considerations in outage plannin PDR O

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DETAILS ,

a 1. Persons Contacted j

• J. Blosser, Plant Manager .l

• H. Bono, Supervisory Engineer, Quality Assurance j

• M. Evans, Superintendent, Health Physics

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• J. Kovar,' Engineer, Quality. Assurance

• R. Miller, Supervisor, Radioactive Waste and Transportation-

• K. Mills, Engineer, Quality Assurance 1

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• J. Neudecker, Supervisor, Health Physics Operations

• E. Olson, Supervisor Chemistry l *A. Passwater, Manager, Licensing and Fuels  ;

E *S. Petzel, Engineer, Quality Assurance L G. Randolph, Vice President Nuclear Operations _

• S. Sampson, Supervisor, Site Licensing )

l *C. Sitzewski, Acting Manager, Quality Assurance ,

D. Schnell, Senior Vice President, Nuclear

• D. Calhoun, Resident Inspector B. Bartlett, Senior Resident inspector

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The inspectors also interviewed other licensee personnel during the

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course of the inspectio * Denotes'those present at the onsite exit meeting on October 8, 1993.

2. General :I l

This inspection was conducted to review aspects of the licensee's l l radiation protection and outage planning programs. Special emphasis was given to those outage planning efforts developed to compensate for the

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effects of the failed fuel detected in May 1993. The inspection included tours of radiation controlled are'as (RCAs), observation of licensee activities, review of representative records, and discussions .I '

with. licensee personne . Plant Tours (IP 83750_).

The inspectors toured outdoor RCAs, the auxiliary and radwaste buildings, and the containment. Overall, housekeeping was very good and radiological postings and barriers were consistent with regulatory and L

procedural requirements. In containment, the inspectors noted poor lighting at the RP contral point at the entrance to inside the i biological shield and the lack of signs indicating low dose waiting .. j area A licensee representative acknowledged those observations and d indicated they would be addressed. Dose rate measurements were also i taken by the inspectors and were in agreement with licensee {

measurements.. During several tours of station RCAs, the inspectors i noted the presence of RP and station upper managemen 'i No violations of NRC requirements were identifie ;

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i l_ Implementation of Outaae Plannina (IP 83750) i As discussed in Inspection Report No. 50-483/93016(DRSS), ALARA planning for the refueling outage had been excellent. Observations during the current inspection, at the start of the outage, indicated that the plans were also well implemented. Minor problems were encountered. and resolved with the timing of a job on the steam generator secondary side and with increased-count times with the personnel contamination monitors +

(because a ventilation system misalignment increased the noble. gas background). In addition, discussions at one of the afternoon outage l- meetings indicated problems with the lack of scaffolding for several

! jobs. Further review, however, showed that the work groups apparently l in need of scaffolding were not adhering to the outage schedul ,

i Discussion by station management at a subsequent outage meeting ,

j emphasized the need to follow the schedul l No violations of NRC requirements were identifie *

l Exposure Control (IP 83750) .

A review by the inspectors of dose rate survey data collected shortly after the outage started indicated that the fuel leak, in-general, did not result in a significant increase in dose rates compared to the previous refueling outage. This was apparently due to the well

. controlled reactor shutdown, early boration and hydrogen peroxide l addition, and subsequent coolant cleanup. Despite these efforts, however, the licensee encountered 0.12 microcuries (4.44 kiloBecquerels)

per milliliter (uCi (kBq)/ml) of cobalt-58 (Co-58) in the reactor cavity .

water compared to about 0.03 uCi (1.11 kBq)/ml in the previous two  !

l refueling outages. This increase resulted in higher dose rates around

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the reactor cavity and on the refueling machine. Discussion with the licensee indicated that actions to minimize the dose from the higher .

rates were being formulate The inspectors also reviewed several initiatives, undertaken during the outage, that have exposure control significance. Several of these are discusse * Core Exit Thermocouple (CET) Connector and Extension Cable Upgrade--This modification will replace existing high maintenance, i single-pin connectors with more durable, multi-pin connectors. In >

addition to the material condition improvement, time and dose savings will also occur during subsequent thermocouple disconnections and reconnections, a routine refueling outage .

activity. The change in the extension cable, while not resulting in a significant dose savings, is expected by the licensee to '

resolve potential equipment qualification concerns (per NRC Information Notice 92-81). The modification is one of the outage's highest dose jobs (estimated 15.52 person-rem (0.1552 person-Sievert)) and is expected to reduce the dose required per outage for thermocouple work from about 1.5 person-rem (0.015 person-Sievert) to less than 0.2 person-rem (0.002 person-Sievert).

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• Steam Generator Channel Head Wash--For the first time this outage, the licensee will be using a vendor's system to decontaminate the _;

steam generator channel heads. The system uses a high-pressure !

water spray (though not high enough to be considered hydrolazing). !

• Tool and Equipment, and Laundry Decontamination Techniques--for the first time this outage, the licensee will be using a tool and equipment decontamination technique based on pelletized carbon dioxide and will augment the typical wet washing of protective i clothing with an ozone-based cleaning technique. Both techniques l are relatively recently-available methods that the licensee expects will be more effective than previously used method ;

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• In-Mast Sipping System on the Refueling Machine--Partially prompted by the recent fuel problems, the licensee will be ;

modifying its refuel machine to allow in-mast sipping of the fuel l assemblies as they are removed from the core and moved to the j transfer canal. This method, which according to the licensee will j be the first used in the Unites States, is in contrast to the current industry method of in-can sippin The in-mast sipping method identifies leaking bundles through the capture and analysis of fission gases that rise into the mast as the fuel assembly is lifted from the core. The fission gas release is driven by the difference in water pressure between the core and fully-retracted inner mas No violations of NRC requirements were identifie . Chemistry Controls (IP 84750) )

The inspectors reviewed the licensee's chemistry controls as required by the failed fuel action plan. Overall, the chemistry controls appeared comprehensive, well thought out, and supportive of health physics activitie Excellent cooperation between the chemistry and operations groups was also note Anticipating that iodine-131 (I-131) levels could increase 50 to 100 times following shutdown, the chemistry and operations groups limited the downpower rate to 10% per hour and increased letdown flow from 75 gallons per minute (gpm) to 120 gpm, to minimize spiking effects and cleanup time, respectively. About four days were estimated before I-131 levels would decline to 0.01 uCi (0.37 kBq) per gram (uCi (kBq)/g), :

allowing for removal of the reactor head under ALARA conditions. The j actual shutdown resulted in I-131 levels increasing by only a factor of !

14 (to 1.12 uCi (41.33 kBq)/g) and only two days needed for recover Although the I-131 cleanup went well, mitigation of Co-58 was mixe After shutdown, Co-58 levels increased to 2.0 uCi (64 kBq)/ml before declining to 0.25 uCi (9.25 kBq)/ml through early boration and hydrogen peroxide addition. However, problems encountered during pressurizer draining resulted in a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> loss of letdown flow causing Co-58 levels to increase to 0.6 uti (22.20 kBq)/ml before again declining. The l licensee had calculated refuel machine dose rates to be about 5-8 millirem (50-80 microsieverts) per hour based on estimated postcleanup j'

Co-58 levels between 0.03 (1.11) and 0.05 uCi (1.85 kBq)/ml . This

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estimate was revised to 14 millirem (140 microsieverts) per hour owing to actual Co-58 levels of about 0.12 uCi (4.44 kBq)/ml. At the end of i the inspection, the licensee was still evaluating why the Co-58 levels had not declined to about 0.03 uti (1.11 kBq)/ml as in the previous two i outage ,

Based on I-131 levels, an estimated 3-8 fuel defects had occurre These defects apparently resulted from corrosion of the fuel cladding j owing to suspected plateout of lithium and boron on the upper portions l of the fuel. This theory was supported by observations of uneven flux l distribution across the fuel element Fuel scrapes will be collected ,

during the outage and analyzed to determine the crud compositio '

Additionally, the licensee consulted with industry experts concerning i appropriate chemistry controls to mitigate further plateout over the j next operating cycl i No violations or deviations were identifie . Effluent Releases (IP 84750)

The inspectors reviewed the effect on effluent releases owing to the leaking fue Although no noticeable increase in fission product activity was evident I in either gaseous or liquid effluents, total gaseous activity increased from 632 Ci (2.34E10 kBq) in 1990 to 996 Ci (3.69E10 kBq) in 1992, owing to the use of contaminated containers during sample collection. In an errata to the effluent report, the licensee reported the true activity ;

as 400 Ci (1.48E10 kBq). Corrective actions were prompt and included l revising plant procedures and providing additional training to health physics technician The inspectors reviewed the licensee's verification concerning the validity of existing accident analyses given the elevated levels of I-131; no problems were identifie Transuranic (TRU) activity had increased in reactor coolant owing to the fuel leakage. Because of the difficulty in detecting TRUs in airborne effluents, the licensee calculated a Derived Air Concentration (DAC)4 factor by correlating TRU to cerium-144 activity in reactor coolant filter This DAC factor was then used to adjust the alarm setpoint~ on the continuous air monitors to account for any increase in TRU level This initiative was considered proactive considering there have been no indications of steam generator tube leakage, and the reactor coolant leakage rate was below 0.1 gp ]

No violations or deviations were identifie . ALARA (IP 83750)

The inspectors reviewed the ALARA program. ALARA suggestions were formally submitted through the suggestion, Occurrence, Solution (SOS) I program. A cost benefit analysis would be performed on the apparent man-rem reduction potential to determine the feasibility of the

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suggestion. No action would be taken on the ALARA suggestion if_the cost benefit analysis was unfavorabl Another aspect of the ALARA program reviewed by the inspectors was the source term reduction program. In an effort to reduce the cobalt-60 (Co-60) source term, the licensee's engineering group reviewed all the valves in the plant that contained stellite and could. contribute to the -

Co-60 source term. Only significant contributors such as throttle ,

valves were determined to be worth early replacement. Throttle valves in the Chemical Volume Control System (CVCS) letdown lines were replaced in a previous outage. During refueling outage 6, two of the three throttle valves for the seal injection system were scheduled for :

replacement with valves containing "norem" material. "Norem" was ,

developed through an Electric Power Research Institute (EPRI) program and the licensee was participating in an EPRI study to determine the effectiveness of the new material. The other valves containing stellite had been identified so that if replacements are ordered, work planning -

would know not to order stellite bearing replacements. The major emphasis of the source term reduction program was on controlling primary chemistry to reduce and filter out corrosion products contributing to the source ter No violations or deviations were identifie ,

9. Exit Interview (IP 83750)

The inspectors met with licensee representatives (denoted in Section 1)

at the conclusion of the inspection on October 8, 1993, to discuss the scope and findings of the inspectio During the exit interview, the inspectors discussed the likely !

informational content of the inspection report with regard to documents .

or processes reviewed by the inspectors during the inspection. Licensee l representatives did not identify any such documents or processes as I proprietary. The following matters were specifically discusse !

, Very good management support for the ALARA planning efforts for the outage demonstrated by routine plant tours and comments made at outage meetings (Sections 3 and 4), Excellent planning and implementation of chemistry control during the shutdown (Section 6). , The licensee's efforts at source term reduction including participating in the EPRI progra I

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