ML050270492

Information Abstracted from NRC Inspection Reports for the Dresden Generation Station, Units 2 and 3, Involving the Unanticipated Release of Tritium And/Or Radioactive Contaminants from the Dresden, Braidwood, and LaSalle Stations Since 199

ML050270492
Person / Time
Site:	Dresden, Braidwood, LaSalle
Issue date:	03/02/2005
From:	NRC/RES/DRAA/RPERWMB
To:
References
FOIA/PA-2006-0115, FOIA/PA-2006-0130
Download: ML050270492 (67)

Text

January 27, 2005

SUBJECT:

Information abstracted from NRC Inspection Reports for the Dresden Generation Station, Units 2 and 3, involving the unanticipated release of tritium and/or radioactive contaminants from the Dresden, Braidwood, and LaSalle Stations since 1990.

List of Materials.

1. Routine Radiological Controls Inspection at Dresden, dated 1993.

2. Notice of Violation (NRC Inspection Report Nos. 10/94008; 237/94008; 249/94008),

dated June 20, 1994.

3. Notice of Violation (NRC Inspection Report Nos. 50-010/94014; 50-237/94014; 50-249/94014), dated August 24, 1994.

4. Notice of Violation (NRC Inspection Report Nos. 50-010/94015; 50-237/94015; 50-249/94015), dated October 27, 1994.

5. NRC Radiation Protection Inspection Reports 50-237/98004(DRS); 50-249/98004(DRS),

dated February 5, 1998.

6. Circ Water Blowdown Line Vacuum Breaker failure due to low stress, high cycle fatigue, resulting in flooding of Owner Controlled property and discharge outside of NPDES approved path, dated December 5, 2000. (Braidwood Nuclear Power Station)

7. Braidwood Nuclear Station Discharge Pipe Vacuum Beaker Leak Dose Assessment, Rev.1, dated February 19, 2001.

8. Questions from Senator Durbin's Staff Regarding the Dresden Station Tritium Leak. (An email correspondence between the NRC and Senator Durbin's staff regarding a 2004 tritium leak at Dresden)

Attachment 6

613-016 A

-Docket No. 50-237 Docket No. 50-249 Commonwealth Edison Company ATTN: Mr. L. 0. DelGeorge, Vice President Nuclear Oversight and Regulatory Services

-,Executive Towers West III 1400 Opus Place, Suite 300 Downers Grove, IL 60515

SUBJECT:

ROUTINE RADIOLOGICAL CONTROLS INSPECTION AT DRESDEN i I r This refers to the routine safety inspection conducted by Messrs. N. Shah and S. K. Orth of this office on May 3 - 7,' 1993. The inspection included a review of authorized activities for your Dresden Nuclear Station, Units 2 and 3. At the conclusion of the inspection, the findings wereldiscussed with those members of your staff identified in the enclosed report.

Areas examined during the inspection are identified in the report. Within these areas, the inspection consisted of a selective examination;of procedures and representative records, observations, and interviews with personnel.

-No violations or deviations were identified during this inspection.

In accordance with 10 CFR 2.790 of the Commission's regulations, a copy of this letter and the enclosed inspection report will be placed in the NRC Public Document Room.

We will gladly discuss any questions you have concerning this inspection.

Sincerely, M. C. Schumacher, Chief - -

Radiological Controls Section.1

Enclosure:

, Inspection Reports..

No. 50-237/93016(DRSS);

No. 50-249/93016(DRSS),

See Attached Distribution

0. Corrosion in Radwaste Storage Tanks (1P 84750)

In two separate events in September 1992, the licensee observed water leaking from the "B" WST and the Unit 2/3 "B" Condensate Storage Tank (CST) which are both in the protected area. The "B" WST leaked in a pipe tunnel that drained to the plant liquid radwaste system. The CST leaked into soil at the perimeter of the tank. Upon discovery, the licensee began to collect and route the leak to the plant radwaste system. Company metallurgists concluded that the leaks were caused by galvanic corrosion of the aluminum tank bottoms. The licensee replaced the bottoms and is planning to drain and inspect the rest of the onsite storage tanks by the end of 1993.

The event was described in the July-December 1992 Semi-Annual Effluent Report where it was estimated that approximately one microcurie total activity, excepting tritium, was released from the tank in 270 gallons of water. Tritium release was estimated at about one millicurie based on the average concentration in radwastedis'charges.- The licensee's offsite dose calculation indicated no impact from this release on down river water users. Tritium activity in the nearest well was basically unchanged from previous years and were at or slightly above the licensee's lower limit of detection (200 pCi/liter).

Although samples of the affected soil were surveyed with a pancake GM detector in a low background area and no activity above background was detected, no isotopic analysis was performed. The samples were returned to the ground. The need to analyze soil samples in such events was discussed at the exit interview and in a telephone discussion on May 21, 1993. The inspectors also pointed out that information regarding such events must be kept in an identified location for ready retrieval at the time of site decommissioning. Licensee representatives stated that soil samples would be isotopically analyzed and that their decommissioning file would be reviewed to ensure its adequacy. Licensee progress on these matters will be reviewed in future inspections (IFIs 50-237/93010-02; 50-249/93016-02).

No violations or deviations were identified.

11. Exit Meeting The inspectors met with licensee representatives (section 1) at the conclusion of the inspection on May 7, 1993, to discuss the scope and findings of the inspection. No documents were identified as proprietary by the licensee, and no violations were identified during this inspection. The following matters were specifically discussed by the inspectors:

• Misaligned filter paper in Unit 1 Chimney SPING (section 2)

'* Procedural deficiencies (sections 4 and 9) 7

tgaEGI9 UNITED STATES 0

04 NUCLEAR REGULATORY COMMISSION REGION III

ŽIlilhIlijK,t 801 WARRENVILLE ROAD LISLE, ILLINOIS 60532-4351

• • • • + A AMS X Aisa tE JUN 2 0 1994 Docket Nos. 50-10; 50-237; 50-249 License Nos. DPR-2; DPR-19; DPR-25 Commonwealth Edison Company ATTN: Mr. M. Lyster Site Vice President Dresden Station 6500 North Dresden Road Morris, IL 60450

SUBJECT:

NOTICE OF VIOLATION (NRC INSPECTION REPORT NOS. 10- __8 /9400S8j 249/94008)

This refers to the inspection conducted by M. Leach, C. Phillips, D. Chyu, M. Kunowski, E. Plettner, J. Smith, and T. Taylor of this office, and by C. Settles of the Illinois Department of Nuclear Safety, on April 12 through May 25, 1994. The inspection included a review of activities authorized for your Dresden Nuclear Station, Units 1, 2, and 3, facility. At the conclusion of the inspection, the findings were discussed with Mr. S. Perry and those members of your staff identified in the enclosed report.

Areas examined during the inspection are identified in the report. Within these areas, the inspection consisted of selective examinations of procedures and representative records, interviews with personnel, and observation of activities in progress. The purpose of the inspection was to determine whether activities authorized by the license were conducted safely and in accordance with RRE-re-qui-rments._

Based on the results of this inspection, certain of your activities appeared to be in violation of NRC requirements, as specified in the enclosed Notice of Violation (Notice). The violations are of concern because there were multiple examples of inadequate procedures and failure to follow procedures which were identified by the NRC. Good procedural guidance and adherence to procedures are essential to safe operation.

We are also concerned about the inordinately high number of personnel contamination events, a problem that also occurred in 1993 during the Unit 2 refueling outage. To date in 1994, the station has recorded almost 3000 such events. Our review during the current inspection identified a lack of strong

• upper management focus on this problem. We do not believe that the problem can be solved through the efforts of only your radiation protection and station laborer groups. In addition to your response to the Notice of Violation, please provide us with a description of your plans to address this problem and a schedule of expected completion dates.

The persistence of the contamination control problem represented a weakness' The licensee's actions to resolve this problem will be reviewed as anInspector Followzppsjem during subsequent

e. Unit I On May 8, 1994, the licensee discovered water issuing from the ground near an outdoor wall of the Unit 1 decant building. The source of the water was determined to be the Unit I contaminated demineralized water storage tank. The jockey pump for the storage tank piping had been put back into service about 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> earlier on May 7. The pump was secured and the section of underground pipe (which had a 3/8 inches hole in the pipewall from corrosion) through which the water had leaked was replaced. An estimated 50,000 gallons of slightly contaminated water had leaked to the ground and into the storm sewer system before the pump was secured. Although the concentration of radioactive material in the water was below the 10 CFR 20 release limits, the leak apparently should have been identified and stopped earlier.

Radwaste shiftly rounds identified a decrease in the tank level on May 7; however, no actions were taken by radwaste' supervision.

The licensee assembled a team to investigate the cause of the failure, evaluate staff performance related to the event, and assess the confidence level in the use of other similar underground pipes. The results of that investigation will be reviewed du,,rqg" future inspection as Inspector Follow-up Item MWN b') )-

On May 10, the licensee drilled four 50-foot deep wells on the south and east sides of the Unit 1 fuel storage building. These wells were drilled in response to an NRC concern about the potential leakage from the unlined spent fuel pool (Inspection Report 50-010/94009(DRSS)). Gamma isotopic analyses of water from the wells identified only naturally occurring radioactive materials, and no Co-60 or Cs-137, both of which were present in the pool. Analyses of the samples identified elevated levels of tritium. The-tritium level's ranged from 2,172 picoCuries/liter (p/Ci) (80 Becquerels/l (Bq/l)) to 51,368 pCi/l (1,900 Bq/l), with the higher values found in the two wells on the east side of the building (in the direction of groundwater flow). Independent analyses of well water samples by the NRC Region III laboratory yielded similar results. The presence of tritium strongly suggested that the pool was leaking. The absence of Co-60 and Cs-137, and the need for makeup to the pool consistent with evaporative losses suggested that the leak was small. The levels found in the wells were a small fraction of the tritium concentration in the pool, about 2.8 million pCi/l (103,600 Bq/l).

The offsite dose consequence of the leak of tritium was negligible. On May 26, two hydrology specialists from the NRC visited the site to review the monitoring wells. The results of that review will be documented in a separate report.

16

No violations or deviation were identified. Two inspector follow-up items were identified regarding contamination control and licensee investigation of Unit 1 spill.

7. Licensee Actions on Previous Inspection Findings (92701, and 92702)

1. TTWE Viol ation lX23103s; Three instances of workers not wearing required dosimeters.- t'o"Wlseling

-and disciplining of the involved workers and training of other workers were completed. The inspectors had not identified a recurrence of the problem. This item is closed. V

/ 21 fT -0 Me nresolved Item-= M4N ORRM4TDFJ: Poor work request documentation. This itemi-ii's7 diUs'cufssed in detail in paragraph 4.c. The inspectors' findings resulted in a non-cited violation. This item is closed.

ENi~rEdE8,9L-nspector Follow-up Iten% 56.3 250(~tiRSS) 0Z2 DRSS)): Contamination control problem. As discussed in paragraph 6.d, the licensee's efforts were not successful. This item is closed.

iOeTZ-dRUInspector Follow-up Item, Moving fuel in three directions simultaneousUy5.-- The li'censee's evaluation was thorough and the decision to maintain the practice of moving fuel in three directions simultaneously was given considerable thought. The inspectors had no further concerns. This item is closed.

No deviations or violations were identified.

8. Licensee Event Reports (LERs) Follow-up (92700)

Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine that reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with technical specifications.

i I ,)r.-rgnx209O_5W Revision l High pressure coolant injection (HPCI) steam piping found outside the Dresden Final Safety Analysis Report design limits due to absence of grout in the HPCI steam line floor penetration. The inspectors reviewed nuclear work request (NWR)

D16705 to verify addition of the grout around the floor drain. The inspectors had no further concerns. This LER is closed.

J Revision, HPCI declared inoperable due to turning gear failure. The inspectors reviewed Dresden Electrical Surveillance (DES) 2300-02, "HPCI Turning Gear Preventive Maintenance" to verify inclusion of the appropriate steps to prevent recurrence. In addition, NWR D10290 for Unit 2 and NWR D19335 for Unit 3 were reviewed 17

i!' -'-< I-if NV%RECU,< UNITED STATES LJ'V .';

-, NUCLEAR REGULATORY COMMISSION 6 0 REGION III

'irir111/S. 801 WARRENVILLE ROAD CAl LISLE. ILLINOIS 60532-4351 A MS AUG 2 1i94 Docket Nos. 50-10; 50-237; 50-249 License Nos. DPR-2; DPR-19; DPR-25 Commonwealth Edison Company ATTN: Mr. J. Stephen Perry Vice President BWR Operations Dresden Station 6500 North Dresden Road Morris, IL 60450

SUBJECT:

NOTICE OF VIOLATION (NRC INSPECTION REPORT NOS. 50-010/94014; 50-237/94014; 50-249/94014)

Dear Mr. Perry:

Enclosed are the results of our inspection conducted by M. Leach and others of this office, and by C. Settles of the Illinois Department of Nuclear Safety, on July 5 through August 15, 1994. The inspectors reviewed activities authorized for your Dresden Nuclear Station, Units 1, 2, and 3. At the conclusion of the inspection, the inspectors discussed.their findings with members-of your staff.

The areas examined during the-inspection.-are identified in the report. Within these areas, the inspection consisted of.selective examinations of procedures and representative records, interviews with personnel, and observation of activities in progress. The purpose of the inspection was to determine whether activities authorized by your licenses were conducted safely and in accordance with NRC requirements.

Overall, we found that the conduct of activities during this period were adequate. However, your performance remains weak in several areas including:

conduct of 'operations, the performance of your high pressure coolant injection (HPCI) system,'and-radiological work practices. In addition, several violations of NRC requirements were identified regarding the following:

• failure to implement proper test controls,

• two examples of failure to follow procedures (unauthorized temporary alterations and radiation protection),

• failure to submit a licensee event report for a failed local leak rate.

test, and

• failure to perform a 10 CFR 50, Appendix J, Type B test on a check valve hinge pin flange.

p .

., Z_-

I 'A F limits were approached; however, the intakes were unplanned and apparently could have been prevented. Review of the licensee's investigation is an Inspector Follow-up Item (50-237/249-94014-11(DRSS)).

4.1.3 Radiation Worker Practices and Updated Survey Maps On August 8 inspectors observed two laborers exit the RPA of the Unit 2 high radiation sampling system building several times without a proper survey. The survey was required by the workers' radiation work permit (RWP) and the workers were required by Dresden Administrative Procedure DAP 12-25, "Radiation Work Permit Program," to follow the RWP.

10 CFR 50, Appendix 3, Criterion V, required that activities affecting quality shall be accomplished in accordance with procedures. Failure to follow the RWP is another example of Violation (50-237/249-94014-07b(DRSS)). In addition, the inspectors noted a discrepancy between the postings of the condensate storage tanks and high radiation sampling system work areas. The licensee corrected the discrepancy.

In addition, the inspectors observed several other poor practices.

These included workers inappropriately leaning into RPAs, workers depositing used protective clothing and trash on the floor of the Unit I turbine building machine shop area, and the use of a bag intended for storage of radioactive material as a makeshift deflector for the exhaust of a small air conditioning unit. These items were discussed with radiation protection department management who indicated actions to rectify the problems would be taken. The inspectors also observed that several survey maps posted on the board for workers to review prior to entering various RPAs had not been updated to reflect current conditions. For example, the posted survey of the Unit 2 517-foot elevation did not show that the drywell hatch was opened and a contaminated area zone had been established at the entrance, and the posted survey for the Unit 3 high pressure coolant injection room did not show that a contaminated area zone had been established at the entrance to that room. The licensee was to evaluate the survey posting practices to ensure that current information was available to the workers. This is an Inspector Follow-up Item (50-237/249-94014-12(DRSS)) pending review of the licensee's corrective actions.

4.1.4 Tritium in Storm Sewers On July 12 the licensee reported to the NRC that recent water samples of storm sewers near the Units 2/3 condensate storage tanks (CST) contained elevated levels of tritium. The highest level was approximately 465,000 picocuries/liter (17.2 x 106 Becquerels). The sampling was conducted as part of an ongoing project to characterize groundwater flow at the site and estimate the amount of leakage from the Unit 1 spent fuel pool (Inspection Report No. 50-237/249-94008). The licensee subsequently identified the "Al CST as a possible source of the tritium and began work to replace the tank bottom. The inspectors noted several other tanks and pipes containing radioactive material leaked or were found to have corrosion (Inspection Report No. 50-237/249-93016(DRSS)) including:

20

/

-I the "B"CST and the "B"waste sample tank which were repaired late in 1992-early 1993, and an outdoor, underground pipe of the Unit 1 contaminated demineralized water system (Inspection Report No. 50-010/237/249-94008), which was repaired in mid-1994 and subsequently abandoned in place. The licensee anticipated the completion of an ongoing cathodic protection modification would reduce such corrosion.

Review of the licensee's evaluation of the groundwater and sewer water tritium is an ongoing Inspector Follow-up Item (50-010/237/249-94011-08(DRSS)). To date, no regulatory limits for discharge of tritium to the unrestricted area have been exceeded.

4.1.5 Radiation Protection Department Staffing During the outage, the inspectors noted that the radiation protection department had a vacant senior manager position and several vacant health physicist positions. These vacancies exacerbated the impact on the radiation protection department of a heavy outage work load and of issues related to the operating Unit 2 and the shutdown Unit 3.

4.2 Security During the inspection period, the inspectors monitored the licensee's security program to ensure that observed actions were being implemented according to the approved security plan.

4.2.1 Incore Detectors Found in Unit 1 Equipment Storage Room On July 14 the licensee found several bags containing twenty incore detectors stored in an equipment room in the Unit 1 containment. The detectors were purchased before 1979 and were never used. The licensee was unaware of the detectors; therefore, the licensee did not account for these detectors in the special nuclear material inventory program.

The licensee immediately initiated corrective action which included searching for more material and properly processing the found material.

Review of the licensee's root cause investigation and corrective actions is an Inspector Follow-up Item (50-010/94014-13(DRSS)).

4.3 Emergency Preparedness and Fire Protection The inspectors verified the operational readiness of the control room, technical support center, and operation support center. Non-routine events were reviewed to insure proper classification and appropriate emergency management involvement. These events are discussed in paragraph 1.2.

4.3.1 Operations Support Center Readiness During Pre-exercise Drill The inspectors observed the licensee's pre-exercise drill on July 13 to evaluate the readiness of the Operations Support Center (OSC). The condition of the OSC was acceptable. Communication at the beginning of the drill was difficult and at times impossible due to severe static in the phone lines. Poor communications later resulted in an operator 21

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-I.- UNITED STATES r

-,°, NUCLEAR REGULATORY COMMISSION C, (tM myf~. oREGION III /

g0 B y IIt"4 801 WARRENVILLE ROAD /

', ae LISLE, ILLINOIS 60532-4351 October 27, 1994 Commonwealth Edison Company ATTN: Mr. J. Stephen Perry Vice President BWR Operations Dresden Station 6500 North Dresden Road Morris, IL 60450

SUBJECT:

NOTICE OF VIOLATION (NRC JNSPECTION REPORT NOS. 5O-O10/94015.7QX 50-237/% 0-249 -r

Dear Mr. Perry:

Enclosed are the results of our inspection conducted by M. Leach and others of this office, and by C. Settles of the Illinois Department of Nuclear Safety, on August 16 through October-5, 1994. The inspectors reviewed activities authorized for your Dresden Nuclear Station, Units 1, 2, and 3. At the conclusion of the inspection, the inspectors discussed their findings with members of your staff.

The areas examined during the inspection are identified in the report. Within these areas, the inspection consisted of selective examinations of procedures and representative records, interviews with personnel, and observation of activities in progress. The purpose of the inspection was to determine whether activities authorized by your licenses were conducted safely and in accordance with NRC requirements.

Overall, we found the conduct of activities during this period were adequate.

We are encouraged by the increased management involvement in daily plant activities and will evaluate the effectiveness in future inspections.

However, during this period, several of your activities were violations of NRC requirements or deviations from previous commitments. In general, these discrepancies involved identification and resolution of past engineering decisions; weak communication and coordination; and inadequate system configuration control. The details of-the violations, deviations, and-non-cited violations are discussed in the enclosed Notices and inspection report.

With regard to the identification and resolution of past engineering decisions, numerous issues emerged including failure to control design changes for motor control centers, inoperable check valve due to a known failure mechanism, and pipe stresses exceeding the American Society of Mechanical Engineers code allowable stress values. Although your staff implemented actions to resolve each specific issue, we are concerned by the potential for other similar conditions and their cumulative effect on plant safety. We request you evaluate these and other examples and determine if any appropriate actions are warranted. Please respond within 60 days to this request with your observations and strategies to this concern.

given little time to plaln and implement the unitization of the department once the decision to do so was announced. The functioning of the re-organized RP department will continue to be reviewed during future inspections.

4.1.5 Tritium in Storm Sewers and Site Cathodic Protection As discussed in Inspection Report 50-010/237/249-94014, the licensee identified leaks in outdoor storage tanks and underground pipes, some containing radioactive liquid. The most recent problem, with the "A" condensate storage tank (CST), was resolved by replacing the-tank bottom.

In the fall of 1993, as part of the review of the leak problem in general, the licensee determined that the existing (distributed or shallow anode) cathodic protection system was degraded. A new system was recently installed, consisting of anodes in 9 deep wells. In addition, a system engineer was assigned, a position previously not filled by onsite personnel. The new cathodic protection system was expected to reduce the corrosion of components in contact with the ground.

Following the "A" CST repair, tritium continued to be identified in water samples from storm sewers in the vicinity of the tank, indicating another source of the tritium. Two pits were subsequently excavated near the tank to allow examination of associated underground pipes. Samples of water that seeped into the pits showed the tritium concentrations in the west pit as high as 800,000-picocuries/liters (29.6 x 10' Becquerels (Bqs)/l) and in the east pit as high as 19,000 picocuries/liters .(0.7 x 106 Bqs/l).

0 The licensee suspected thatthe tritium source in the west pit was possibly the 18" HPCI test return line. On October 3 the licensee began pumping water from the east pit to a storm sewer that discharged to the Unit 2/3 discharge canal, upstream of the normal liquid radwaste discharge point. Pumping was necessary because water seepage into the pit prevented examination of the pipes for leaks. Periodic grab samples were taken and analyzed by the licensee for tritium and gamma-emitting isotopes because the storm sewer was not monitored. The licensee's calculations showed the offsite dose from the tritium in this water would be well below reguliatory limits. This is considered an Inspection Followup Ite7 OI V* ) pending further review of the licensee's 10 CFR 50.59 evaluation

- the release of the water through the storm sewer.

4.1.6 Unit 1 Activities On August 31 an operator identified a level decrease of about 1.5' in the Unit 1 spent fuel and transfer pools. An investigation was initiated to determine the cause of the decrease, including a walkdown of the accessible areas of the Unit 1 sphere to look for leaks. The licensee concluded that the decrease resulted from a resin transfer from the pools' temporary demineralizer system to a radwaste liner (the resin was being changed because of a microbiological growth problem in the pool). About 1000 gallons of pool water was used in this transfer. The workers transferring the resin failed to communicate to the operations staff the effect the transfer would have on pool level. The licensee took actions to ensure future evolutions involving the pool would be communicated to 19

February 5, 1998 Mr. Oliver D. Kingsley President, Nuclear Generation Group Commonwealth Edison Company ATTN: Regulatory Services Executive Towers West IlIl 1400 Opus Place, Suite 500 Downers Grove, IL 60515

SUBJECT:

NRC RADIATION PROTECTION INSPECTION REPORTS 50-237198004(DRS); 50-249198004(DRS)

Dear Mr. Kingsley On January 21, 1998, the NRC completed an inspection at your Dresden Generating Station, Units 2 and 3. The results of the inspection were discussed with Mr. Larry Aldrich and other members of your staff on January 21, 1998. The enclosed report presents the results of this inspection.

The purpose of the inspection was to review the unconditional material release program, and the effectiveness of the actions taken on previous open items and violations.

Overall, the unconditional material release program was comprehensive, technically sound and well implemented. Actions taken to prevent recurrence of previously identified violations, inspection follow-up items, and licensee event report items appeared effective.

No violations of NRC requirements were identified during the course of this inspection.

In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter and its enclosures will be placed in the NRC Public Document Room.

R8.3 (Closed) LER No. 94-025-00: Inadvertent placement of the Units 2/3 main chimney monitor (SPING) into the flush mode by a chemistry technician during a routine sample change. The problem could not be immediately corrected so operations switched to the backup (GE) monitoring system. However, when operations switched over they discovered difficulties in placing the GE system into service because of deficiencies and ambiguities in the operation procedure DOP 1700-11, and chemistry procedure DCP 2213-01. To correct this problem the licensee revised procedure DCP 2213-01 to include in the checklist an initialed step and verification that the SPING was found operable before changing the samples. In addition, operation procedure DOP 1700-11 was enhanced to provide clear instructions on the method for switching systems following flow oscillations. The inspector reviewed these revisions and noted they had been accomplished.

R8.4 (Closed) Violation 50-237/96009-1 0: 50-249/96009-10: Failure to perform an adequate survey to assure compliance with 10 CFR 20.1201 (a)(1)(1) which limits radiation exposure to the Total Effective Dose Equivalent of 5 rems per year. The licensee failed to properly evaluate the potential radiological hazards, and use process or other engineering controls to control airborne radioactivity concentrations during the removal and transfer of contaminated bags of radioactive material. The corrective actions to prevent recurrence included revision of procedure DAP 12-09 to require a radiation protection shift supervisor at prejob briefings for work identified as high risk, reiteration of expectations to radiation protection technicians regarding the necessity of the duty supervisor to be informed of all work activities and job scope changes, presented a training course to radiation protection technicians entitled "Conservative Decision Making", and discussed this event, radiation work permits, and expectations concerning v survey requirements.

R8.5 (Closed) IFI 50-237/94-015-07

50-249-94015-07: Evaluation of water with tritium concentrations released through the station storm sewer system. On October 3, 1993, the licensee pumped contaminated water (tritium) to a storm sewer that discharged into the Unit 2/3 discharge canal. Periodic grab samples were taken and analyzed by the licensee for tritium and gamma-emitting isotopes because the storm sewer was not monitored, nor was it the normal handling system for discharge. Sample analysis indicated that offsite dose to the public was well below regulatory limits. In accordance with technical specification requirements, the licensee issued an LER for this event because the liquid radioactive waste was not processed through the normal waste handling system. The discharge was incorporated into the Station Semi-Annual Effluent Report for the period of July-December, 1994. In the thirty day report issued to the NRC, dated December 9, 1994, the licensee listed actions taken to prevent recurrence of this type release. The inspector reviewed the actions and concluded they were implemented.

V. Management Meetings X1 Exit Meeting Summary On January 21, 1998, the inspector presented the preliminary inspection results to licensee management. The licensee acknowledged the findings presented at the exit meeting.

The licensee did not identify any information discussed as proprietary.

5

I TITLE: Circ Water Blowdown Line Vacuum Breaker failure due to low stress, high cycle fatigue, resulting in flooding of Owner Controlled property and discharge outside of NPDES approved path.

UNIT: Braidwood Station Unit Common EVENT DATE: 11/06/2000 EVENT TIME: 14:30 REPORT NUMBER: AT # 3 8237 /CR #A2000-042 81 REPORT DATE: 12/05/00 REVISION: 2 INVESTIGATORS: Mike Riegel (Team Lead), Paul Uremovic (Qualified Root Cause Investigator), Luis Rhoden, Kimberly Aleshire, Joe Tidmore, and Harry King ABSTRACT:

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.,$ . .w .-. .. . s thfoevintis notf hitsverexceedddlirefore 'a The leaking valve was replaced and h; illdt I be y ibwapupdackI into the CWTBlowdown pipimgxA root cause investigation was conducted to deteruine factors that contrfibuted to the failure of the valve and subsequent release of water. Analytical techniques employed in peiforming the root cause investigation included Failure Modes and Effects Analysis, Barrier Analysis, Event and Causal Factor Cha ring and interviewing.

The vacuum breaker valve failed due to low stress high cycle fatigue. This valve had been in service since initial construction. Root causes for theyvalve failure were lack of ani adequate preventive maintenance program for the vacuum breaker valves and an inappropriate configuration (lack of internal surge protection) of the currenly installed vacuum breaker valves.

Additionally, system operating methodology is a significant contributing factof in that the operating procedure allows reinitiating blowdown flow following short duration shutdowns by using the motor operator on the isolation valves. Opening valves using the motor operator

-he rapidly establishes full flow and causes significant pressure surges since the piping will be depressurized and partially drained.

f%7 17 Inlece6rive "to^" 5 V go I I tA %All-

The failure mechanism and root causes would be applicable to all vacuum breaker valves in the Circ Water Blowdown and Makeup (MIU) systems. Similar systems are installed at Byron and LaSalle stations. However, Byron uses a different type valve.

Corrective Actions include implementing adequate preventive maintenance programs, replacing the valves and revising the operating procedure to prevent power opening of the isolation valves to establish blowdown flow.

CONDITION STATEMENT:

The float assembly for the Circ Water Blowdown Vacuum Breaker, OCW136, experienced fatigue failure due to impact loading from excessive operating cycles and inadequate pre'entive maintenance. Gosrbyjsn~euapoe 8!

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tU;4iEl7iilg1weiiiiw E7T~2m~ithinii sigall iff

_____ Ex'ewatof"unresro cted drik sgou~e 'lidits'.T h sdi-ctvey-ongighfyin-R et i e radioact~iity'samnple results mi~d Lower Limiits of Detection:

-_(LLIJg are in Attachmzent 6. provided maximum extent feasible, all water was pumped 'back into the>CW~Blowd 1iie.

EVENT DESCRIPTION:

(Refer to detailed timelinie (Attachment i) for additional iniformation.)

tiiindiiffedin aas e itrerW peF 89h7§ .sfid h fetad h toavlrie j-th e dit'chwaii'siilaeyresimatd.Supctn that a faulty vacuum to b aprxnitl allo on 80O~breaker either the Cire Water M/T or Blowdown Sstem was-the source of the water, the NPDES Coordinator notified the Shift Manager and 0CC Director of the IEPA notification.

EVENDEDEIPION ' ' :. . o6-~2- -n ';

ati approximoately 15:00 The NPDES aeS At approD Coordinator notfie walked down the Cire Water thIP~f~

IWOdinto fit"D h systemm'izn Blowdown sourc S- and We! -

and identified that the water was coming from a valve vault that houses the 0CW13516 vacuum breaker/isolation valve assembly. The NPDES Coordiniator assessed the site Dand concluded that the water was confied to Site property anid the- 6dL

~~~~A, ire immediately adjacent to Site property.

A

==_ ditchtbarea along the south sidfof Srniey Road t-°;i'5go

• slt'44. oK;e

'Q The water in the ditch was confined by the resident's drveway to the west and by higherD -

elevation to the east: chrn iittertiv EAjconcludcd thatRn dd l I

' 1i  ;-syequired and.ithat 4here wre no NPDES;Afi

.eto_ n Statet.

2 X e ClnCIst cIU atU or Idv o ernnw war tl lE

Between 16:00 -17:00, a meeting was held with senior station management, the Shift manager and the OCC staff. The NPDES coordinator briefed the attendees on the results of his field observations of the area surrounding the vacuum breaker valve. Station nanagement was also briefed on the discussions between the NPDES coordinator and the [EPA. Senior management directed the following actions be taken:

1. Operating personnel were to evaluate water inventories and to explore potential alternate release options.

2. Isolate the CW Blowdown System

3. Make preparations to take the C(W blowdown system out of service, drain the piping section and replace the failed vacuum breaker valve.

The Circ Water Blowdown system was then isolated in preparation for draining and repairs.

There was no discussion at this time of any need to sample for radioactivity in the water that had been discharged.

At approximately 0615 on 11/n/00, the RP Manager was contacted by the Operations Manager that there was a blowdown line leak and that RP was requested to meet with 'the Chemistry Manager'to look at potential alternate release paths for radwaste. The reason for this request was that radwaste releases would not be possible via the blowdown system while blowdown was isolated for repairs to the vacuum breaker valve.

Following this phone conversation, the RP Manager spoke with the RP Technical Superintendent and discussed the need to pull water samples. Included in this discussion was a conclusion that the samples should contain no radioactivity because of the belief that the spilled water was "only lake water".

A decision to conduct water sampling of vacuum breaker structure was made at 0800. The sample was taken-at approximately 0845 and the results of the gamma isotopic analysis indicated no 'quantifiable peaks found (NQPF), and the tris iithe lower limit ofdetectioniiLLDLof At approximately 1130, RP received information that the leak may have occurred for a period of 7-10 days and that the water that leaked was from the'circulating water blowdown line which carries the liquid radwaste discharges from the station'to wtheriver. At 1230 on I101100, a decision was made to initiate soil sampling in the vicinity of the vacuum breaker structure, and to obtain a water sample from the standing water that was onsite, but near the Smiley Road ditch.

At approximately 0830, Mechanical Maintenance (MMD) personnel with assistance from System Engineering pumped outthe'OCW135/136 vault and began draining the blowdown piping to facilitate work on 0CW1351136.AAt approximately 1200, after the CW Blowdown line had drained sufficiently, the entire OCW 135/136 isolation valve and vacuum breaker assembly was replaced.

3 a.. . - -.. 0 ~. .. r--*b, 9**. at--lf

A total of 5 soil samples were obtained within approximately 30 feet of the vacuum breaker structure, and 2 of the 5 samples had detectable levels of radioactivity. 4heffsf itl, tbtained~ near~ Raodtlgiiiiislpc

~ ~ ~ ~ d~fiegamA rholey indctd~2Ernd vah~'ter-analysis iindicatd trxtnum~at a level ~of.3SE-5 uCi/mli-)

The results of these samples were discussed with corporate Generation Support Department (GSD) RP Manager at 1900 on 11-7-00. Corporate GSD agreed to discuss the issue with the corporate Generation Support General Manager. At 1945, the Station Manager and Site Vice President were notified of the sample results.

At 0830 on 11-8-00, the RPM discussed the sample results on the morning call. At 1400, the RP, Chemisty, Regulatory Assurance, Station Manager, and Site VP met to discuss the current status, next steps, and sampling for the event.

£Rn,1'l-@04ditio at-lsite-ai HI~xig21ealLteLgRiaa-.sml atS~jamp~wereW& taken -at~approbxirn-&ely!4600:pndreut indiLca id~rtuffievels3-7jF054 to 3E79.5,Ci /ml. f^S*S On 11-9-00 at 1000, a conference call was held with the site and corporate and an Offsite Sampling Plan, Renediation Plan, and Communications Plan was agreed to. At 1200, discussions were held with site NRC aid regional NRC. At 1210, nbtification of the offsite release was made to Will County authorities and to the Reed Township -Highway Commissioner.

At 1245, RP was dispatched to obtain water samples from the Smiley Road ditch.

The etpvprninAtolyf 1-400, teuRC Regioaml Oie aobnaMdD wr ntdthe SmileyfR6AdoZitoW Gadma' t(S lsesamle~wee aso ayzeddby4Tefled' "~Isotope§Mid*&estaiebootidr-y.'W-The evening' of 11-9-00, the NRC Regional Office and ]DNS were notified of the Smiley Road ditch sample analyses results.

At approximately 1100 on 11-10-00, pumping of the water back to the blowdown line commenced. Pumiping continued using a 600 gpm pump, approximately 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> per day, until 2000 on 11-15-00 when all possible water had been pumped back into the blowdown line.

4

BACKGROUND INFORMATION:

The primary function of the Circ Water Blowdown System is to provide for Lake turnover to prevent undesirable chemical buildup in Lake. The secondary function of the Circ Water Blowdown System is to provide dilution for liquid releases.

The Circ Water Blowdown System (Attachment 7) is designed to return Cooling Lake water back to the Kankakee River. Processed fluids from the Sewage Treatment System afid the Radwaste Treatment Systems discharge directly to the Circ Water Blowdown system where dilution occurs prior to release to the Kankakee River. The Wastewater Treatment Plant and the Dernineralizer Regenerant Waste systems along with various strainer/filter backwashes are returned to the Cooling Lake and thus are indirectly returned to the Kankakee River through the Blowdown line after dilution by the Cooling Lake.

The Circ Water Blowdown system begins at the Circ Water System supply to the condenser.

ntuitIfd e a d18 onbi on seacf2eVTr'e'.tiyee di dsarge'pip'e connectstoWh t -

Z36 Bldii~rj6 conectio of thenecirftb'1 _' ~ ad BloexaoWn ai pipe iniiipeR the fly w tr l 7tL high f ppelran Uon pipevacuumpein breakrisroil ropaas pital functionis d to treleasecenraiepipr'P tatprnthepKatthe high p dusing normal pio aleasofcltates8 edarlertpnngo h vaumn brekeon. syThm&shutRown- On seplfot v shutowgnsalqirpocets peration.~wtha Udevlop se n athownth hihpit' d e vacuum ea b oiiv reaker is tos open rsue a4s edn ohl axgedps txheacuum bIae nissa vn thug waerlvlisbelo thvlaseby However, thea piotvalvem wil reesUh i n llwtevcu ~ to rae open as sona evldos E-typopach Epheauuvacuumreakto breakerpreddwts is srown on ensu atterl utrlis ison ioainvlets on sytem eer aiitte vacsutmbw vcu broenake eker shutns mainenanveair dcr.procthatpdevelop ss h i ~esat rpltxlepoieg pots may b positivpessre, w unc'os'hpimr eni to holdio vavfntinist rlas 5ra ntan-ed ar htacuu-ltesatte ih"Oin', uin

a. . . -

The Circ Water Blowdown system was originally designed to be maintained full and pressurized.

This was accomplished through manipulation of the Blowdown Spray Valves, OCW018AlB.

These valves were susceptible to freezing due to design and system operation requirements.

Based on this, other maintenance issues, and parts obsolescence, these valves were eventually abandoned in the full open position in the mid 1980s. System control was transferred to the upstream motor operator isolation valves, 1/2CW018. The system would no longer be maintained full and pressurized on shutdown as 'a result of this change. Minimal technical review was performed on the hydraulic effects on the vacuum breakers from this method of operation (ie: surge check valves were not evaluated for system incorporation).

In 1997, Chemical Feed System was relocated from the Turbine Building to the Lake Screen House under Modification M20-0-95-003. One of the primary reasons for centralization of the Chemical Feed system to the Lake Screen House was to reduce maintenance cost via system size. This design change necessitated isolating the Circ Water Blowdown System on a daily basis to accommodate biocide injections into the Circ Water System, because our permits do not authorize discharge of biocide to the Kankakee River.

The daily requirement to isolate Circ Water Blowdown for biocide injection prompted the Operations Department to challenge the UBwOP CW-12 procediiral requirement to slowly open the l/2CW018 valves for system start-up.' BwOP CW12 was revised to allow fast motorized operation of 1t2CW018, in lieu of slower manual throttling, following short periods of system shutdown (i.e.: biocide injections). ' Minimal technical review was performed on the hydraulic effects on the vacuum breakers from this method of operation.

Work history on the Circ Water Blowdown System vacuiiuim breakers was reviewed. There were no recorded vacuum breaker float asserimbly failures prior'to this event. Several instanices of

'leaking pilot valves were noted from the review.- The OCW060 pilot valve was discovered leaking in 12/98. PIF # A1998-04324 was generated to address the flooding of site property and the Smiley Rd ditch immediately adjacent to site property. The piping to 'the air'release valve on the 0CW058 failed in 12/96. The complete vacuum breaker assembly including pilot valve was replaced with a new assembly in 6/97. It should be noted 'thatte 0CW 058 vacuum breaker failed again on-about 11/20/2000 while this root cause investigation was in progress. The float assembly broke at the bowl to guide bar weld. No other significant work history was identified.

The failure of the OCW136 float assembly was discussed with the vendor. Based on the failure description, the vendor indicated that it appeared to be consistent with the effects of a pressure surge (i.e.: water hammer). The vendor indicated that surge protection check valves should be considered for a vacuum breaker when pipe flows exceed 6 ftWs and are required when flow velocities exceed 10 ft/s. The vendor also recommended a 7-10 year PM frequency to address valve elastomer degradation.

6 MT# 1,,10-0101 "- I, I , II U r

ROOT CAUSE ANALYSIS AND CORRECTIVE ACTIONS:

A. Investigation and Root Cause Analysis Techniques Methods utilized for performing this investigation included interviewing, Barrier Analysis, Event and Causal Factor charting (Attachment 2), and Failure Modes and Effects Analysis (Attachment 3). Task analysis had also' been specified in the root cause report charter. However, since task analysis is a tool that is used on investigations where problems during performance of tasks contribute to the event and no such performance issues were identified during this investigation, the task analysis technique was not utilized. The majority of the conclusions in this report are based upon outcome of the Failure Modes and Effects Analysis and Barrier Analysis. Event and Causal Factor charting did lead to the development of some questions, and the identification of potential barriers, but otherwise did not provide much useful assistance in this investigation.

Root Cause Analysis The investigation determined that inadequate material condition of the OCW136 Vacuum Breaker Valve resulting from inadequate preventive 'maintenance, inappropriate configuration of the vacuum breaker valve assembly and system operation methodology led-to valve' failure and release of water from the blowdown system.

' Event and Causal Factor Chart

.An Event and Causal Factor Chart (E&CF) is included as Attachment 2. As stated above, the E&CF chart was used principally to identify potential barriers and to develop questions for interviews, 3Barrier Analysis Barrier analysis shows 3 barriers which could have prevented or mitigated the vacuum breaker valve failure. A discussion of each barrier and its failure method/mode is provided below.

Interview results contributed significantly to the barrier analysis A. Preventive Maintenance - This program ensures that equipment and systems are checked periodically and maintained within acceptable parameters'so they will function as designed. The preventive maintenanceprogramhas no requirement to perform any ind of internal valve inspection or operationalcheck and no requirementto periodicallyreplace-the.valves.

The vacuum breaker valves were-essentially installed as run to failure components.

aThere re no Technical tion requirenents or NRC commitments -to conduct periodic mnaintenance. Prior to 1999, informal walkdowns of the blowdown system were performed on an annual basis.-'For tho most part, results of the walkdowns were not logged and no records maintained of the observed materal conditioi of the valves, other than action requests for repair of observed deficiencies (X.l.'ek ing pilot valve). A handwritten record of a 1995 walkdown of both the CW Blowdown and Makeup lines was found. This record noted whether valves were open or shut, any leaks found, and any water in the valve vaults.

7

In July 1999, a preventive maintenance template from STANDARD NES-G-08, ComEd Performance Centered Maintenance (PCM) Templates, was adopted for application to the vacuum breaker valves. The particular template chosen is specifically applicable to spring actuated safety relief valves, and contains no discussion of applicability to float type valves. The predefine task description is "sperform setpoint verification and seat leak check, or replace valve". The periodicity was set at 10 years. Although most of the valves had been in service since initial construction and had no previous maintenance history, due dates for maintenance were set well into the future, and no consideration given to replacing any of the valves based upon age or time in service. This issue is also discussed under Root Cause #1 The template chosen was the closest match from all those available in the standard PCM template index. Time pressure to complete the project was given as the reason for choosing this default as o pposed to developing an appropriate template applicable to the vacuum breakers. CA 1, CA 2 and CAPR 1 have been'generated to correct specific maintenance issues associated with the vacuum breakers. CA 4 has been generated to review a sample of preventive main'tenance templates to look for' additional instances of incorrectly assigned te nplates.

B. Design/Auplication - Components utilized in plant construction should be of appropriate' design for a given application to ensure acceptable service. The baierwas challenged when system operation was changed without changing the design or configuration of the vacuum breakerassemblies.

Original CW blowdown system operation provided for controlling blowdown flow using valves at the river screen house, thus the system would'always reImain full of water. This method of operation was abandoned within the first two years of operation due to repetitive failures of the control valves.- Current operation of the system provides for controlling blowdown flow using valves'located in the plant near the main condensers and when flow is'secured, the blowdown line wi depressurize.

and partially drain resulting in a potential pressure surge when flow is reinitiated.

Discussion with the valve manufacturer revealed that if the valves are'subjected to.

significant pressure surges, they should be equipped with surge protection. The current configuration has no'surge protection.; This issue is discussed more fully in the section titlled Root Cause #2.

The reason whylthe system operation was changed rather than correcting the material condition of the valves at the river screen house will not beepursued since that decision was made so long ago. Similarly, the reasonuthe change-was made without considering impact on the vacuum breaker designlconfiguration cannot be determined. CAPR 2 has' been generated 'to replace the current 'design vacuum breaker assembly with assurge-protected configuration.

8

In developing corrective actions, consideration was given to replacing the vacuum breakers with a totally passive standpipe system. However, some of the vacuum breakers are under pressure and the resulting stagnant column of water in the associated standpipe would necessitate installation and maintenance of heat tracing to prevent freezing in winter weather. Therefore, this potential solution will not be pursued any further.

Additionally, consideration was given to restoring control to the OCWO I8A/B valves at the river screen house. These valves are abandoned in the full open position, which reduces head loss and provides for maximum flow. During operation currently, most of the blowdown piping is not water solid and the valves would have to be throttled down to maintain the system full, which would also reduce blowdown flow. Reduced blowdown flow is not desirable from a cooling lake chemistry control standpoint.

The valves could be used to bottle up the blowdow'n system when shutting it down.

However, this would require closing the valves prior to securing blowdown flow in order to fill the piping, potentially resulting in damaging water hammer to the system.

Therefore, this potential solution will also not be pursued any further.

C. Operatine Procedures - Procedures provide the appropriate instructions to ensure actions are' carried out correctly with'appropriate limits and acceptance criteria. This barrierfailedwhen the procedure was-nod ed to allow operathon of the system that could result in significantpressure sures(fatigue cycles) on the vacuum breaker valves.

The CW Blowdown System Startup, Operation and Shutdown procedure, BwOP CW-12 allows opening the blowdown isolation valves with power following "short duration shutdowns" such as routine daily chemical biocide additions. However, during a typical 2-hour shutdown of the system for biocide injection, the blowdown system can partially drain and thus a significant surge (water hammer) can result when flow is rapidly restored using the motor operators. This issue is discussed more fully under the section titled Significant Contributing Cause.

CA 3 has been developed to change the blowdown procedure to always require opening the isolation valves in stages overa few minutes to prevent water hammer.

Failure Modes and Effects Ainalysis This technique was used to develop and analyze potential failure modes for the failed vacuum breaker assembly. A fishbone diagram used in this analysis' is provided as Attachment 3.

Brittle Failure. This mode was eliminated due to the valve float being constructed of Stainless Steel and not susceptible to brittle failure in the system operating environment.

.y Manufacture. This mode was eliminated due to lack of any noted manufacturing defects in the failed valve.

9

Preventive Maintenance. Failure to perform adequate preventive maintenance certainly contributed to failure of the valve, and in fact lack of an adequate preventive maintenance program is one of the root causes. However, there is no evidence that any preventive maintenance ever performed on the valve contributed to the failure, therefore improper performance of preventive maintenance was eliminated as a failure mode.

Corrective Maintenance. Similarly, lack of any significant corrective maintenance history rules this out as a failure mode.

Aging. The failed valve was examined closely and did not show signs of stress corrosion cracking or other corrosion mechanisms. There was evidence of age related wear in the valve bushings, however this is not considered a contributor to failure. The length of service does come into consideration when coupled with the fatigue mechanism.

Design. The lack of surge protection, given the way the system is operated, is a strong contributor to valve failure. This information was provided via interview with the valve manufacturer. Beyond that, the valve design, specifically that of a float operated vacuum breaker, is an appropriate application for'protection of the CW blowdown system'.

Fatigue. Post failure analysis shows that the most probable failure mechanism was low stress high cycle fatigue. Visual inspection reveals a large fatigue and small fracture cross section, con sistentwith'this failure mechanism.

System Operation. This mode is a strong contributor to failure, due to the fatigue cycles caused by operation. System operation is discussed in great detail in other sections of the report and therefore will not be elaborated on in this section.

B. Summary of Causes and Corrective Actions Root Cause #1 The preventive maintenance program for the CW Blowdown vacuum breaker valves is inadequate. Thie valves, with one exception, had all been in service since initial construction of the plant prior to the failure and subsequent replacement of the OCW136 Valve. The other valve replacement was due to problems with the pilot valve sensing line failing (due to general corrosion), and not with the mainvacuum breaker valve itself. Preventive Mainterance essentially consisted of an annual walkdown of the valves by the system engineer, and if no leakage was observed, the valves were deemed to be OK; results of the walkdowns were not logged.

A preventive maintenance template for he valves was adopted from Performance Centered Maintenance (PCM) standard templates in July;1999. Thfie'template chosen was for "Spring Actuated Pressure Relief Valves" which was the closest match fro'm the choices available. The "why" 'givenfor this was time pressure to complete the prIeventive mainmtenance' templates. This cause will not be pursued further because the project is completed and therefore no relevant CAPR would be developed. CA 4 has been generated to sample for additional inappropriate preventive maintenance templates.

10

...- P PatI.. , - . , n.. If.-tt--- - - Sl .

- .111fil

CA 1 has been generated to replace all vacuum breaker valves and thus restore current material condition CA 2 has been generated to develop an adequate preventive maintenance program for the valves which includes periodic inspection of the valves including internals or provides for'valve replacement at appropriate intervals. This item will also include the development of system walkdown inspection requirements including specified frequency of walkdowns and documentation/reporting of walkdown results.

CAPR 1 has been generated to implement the revised preventive maintenance program and system walkdown inspection requirements.

Root Cause #2 The configuration of the current vacuum breaker valve assembly is inappropriate. The current valve assembly (Attachment 4) consists of an' integral butterfly isolation valve and vacuum breaker float valve with an attached pilot.or "air release" valve. Discussion with the valve manufacturer revealed that the current configuration would be susceptible'to premature failure if the valves were subjected to significant repetitive pressure surges during operation, such as would be experienced during startup with rapid filling/pressuriiation of.the system. The manufacturer stated that with pressure surge conditions, a valve with built in surge protection (Attachment 5) would be required.

The original system operation provided for controlling blowdown flow using valves at the river screen house, thus the system would alwaiys remain full of water. This method of operating the system was abandoned within the first two years of operation due to repetitive failures (caused principally by freezing) of the control valves at the river screen ho use, and difficulty, in 'remotely controlling the valves at the river screen house. Current operation of the system provides for controlling blowdown flow using valves located in' the plant nearthe main condensers, and thus when flow is secured, the blowdown line will depressurize and at least partially drain resulting in a potential pressure surge when flow is re initiated. The reason for why system operation waas changed as opposed to correcting the material condition of the valves at the river screen house is beyond the scope of this investigation due to the large time interval from initial construction and will not be evaluated further.

CAPR 2 has been generated to replace the current design vacuum breaker assembly with a surge protected (Attachment 5) configuration. This action can be, erfonrmed coincident with CA 1,but is listed separately to provide a definite link of CAPRs to root causes.

11

Significant Contributing Cause A significant contributing cause, if in fact not a third root cause, is system operating methodology. As stated earlier, blowdown flow is controlled using'valves in the plant located near the condensers, specifically the'1/2CW0I8 valves, which are 24-inch motor, operated butterfly valves. If blowdown flow is initiated by opening these valves using the motor operators, a significant pressure surge can result.

The Chemical Feed System configuration was modified in 1997, changing the injection point where chemicals are added to the CW system, and now CW blowdown flow must be secured whenever chlorination of the CW system is accomplished. This is done to comply with NPDES permits, which forbid discharging of biocide to the river. Chlorination is performed for each unit on a daily basis and therefore it is necessary to alter blowdown flow on a daily basis to ensure there is no blowdown from the unit being chlorinated.

The CW Blowdown System Startup, Operation and Shutdown procedure, BwOP CW-12, contains a note that states' in part "Slowly opening'MOV CW018, Units 1 & 2 CW Blowdown Isol Vlv, in stages oover a few minutes will prevent damage to the CW Blowdown piping/components due to water hammering of the draiwed piping." To provide some relief to the operators from having to manually operate the isolation valves on a daily basis in-support of chorination operations, the procedure was changed in 1998 to also state' For short dtion shutdowns of CW blowdown (i.e.: routine dailyhmical biocide additions, etc.), reopening with power from MCR is acceptable." This change was authorized without sufficientee tchical analysis, and in fact following a typical two hour shutdown for chlorination there is a significant surge whien flow is restored, paricularly if only sin'gle uinit blowdown is in operation and blowdown is totally secured (as opposed to shifting to the oppos ite unit) for chlorination.

CA 3 has been developed to change the blowdown procedure, BwOP CW 12, to always require slowly opening the isolation valves in stages over a few minutes when initiating flow from a no blowdown flow condition. This CA does not apply when shifting blowdown from one unit to another as long as blowdown is not secured.

C. Equipment Failures (EF)

EF# Summary of EF Associated Causal Factor

1. OCW136 vacuum breaker float assembly failure (date of 1,2, and 3 failure estimated to be between 10/27/2000 and 10/30/2000)

2. OCW058 vacuum breaker float assembly failure (date of 1,2, and 3 failure estimated to be 11/20/2000)

Both the failed CW blowdown valves, OCW136 and OCW058 (failed 11/20) were analyzed to determine failure mode, and both were determined to have failed from low stress high cycle fatigue. This type of failure mode is consistent with the causes listed above.

12 n7. It, I J`n1VDC*CI0I DC QMn I a JO It I IJ 10J LI on

D. Causal Factors (CFs)

(Root Causes are identified by Asterisks)

CF # Summary of Causal Factors Associated Correictive Action 1* Root Cause - The preventive maintenance programs for CAPRI, CAl, th e'CW Make-Up and Blowdown Syst'em vacuum breakers and CA2 are inadequate. An effective preventive maintenance program needs to be developed and implemented for both systems.

2* Root Cause - The design configuration of the current CAPR2 and CAl vacuum breaker valve assembly is inappropriate. The current valve assembly (Attachment 4i) consists of an integral butterfly isolation vaive and vacuum breaker float valve with an attached pilot or "air release" valve.

Discussion with the valve manufacturer revealed that the current configuration would be susceptibleto premature failure if the valves were subjected to si-nificant repetitive pressure surges during operation, suchnas would be experienced during startup With rapid filling/iressurization of the system. The nanufacuier stated that with pressure surge conditions, a valve with built in surge protection (Attachment 5) would be required.

3 Significant Contributing Cause - Current system operating *CA3 procedure (BwOP CW-12) allows initiatingr CW blowdown flow by opening the isolation valves using the motor operators if blowdown flow has only been shutdown for a short duration (i.e. routine daily chemical biocide additions, etc.). Operating the blowdown system in thisrmainner can result in significant pressure surges.,

4 Contributing Cause - In July 1999, a preventive CA2 and CA4 maintenance template from STANDARD NES-G-08, CormEd Performance Centered Maintenance (PCM)

Templates, was 'adopted for application to the'Vacuum breake valves. The p articular tmplate chosn' is specifically applicable to spring actuated safety relief valves, and contains no discussion of applicability to float type valves. Time pressure'to complete the project was given as the reason for choosing this default as opposed to developing an appropriate template applicable to the vacuum breakers.

13

E. Corrective Actions (Corrective Actions to Prevent Recurrence are labeled CAPRs)

Immediate Corrective Actions: *

• The CW blowdown system was isolated to stop the leak

• A team was assembled to recover from the leak and restore the system

• Appropriate notifications were made

- The leaking vacuum breaker valve was replaced

• Spilled water was pumped back into the CW Blowdown System

• Radioactivity samples were taken to comply with 10CFR50.75(G).

• A root cause investigation was commenced

• These are all listed as immediate actions even though some occurred over several days.

Actions Corrective Action Assignee -Due Date CAPR 1 Description of corrective action to be A89301T 03/01/01 taken:

implement a revised preventive mainitenance program for the float operated vacuum breaker valve assemblies for the CW Blowdown and Makeup Systems. This PM will be devloped byCA 2 and will include

,, ,if.:It,.S"'-

specific intervals for inspection of valve internals or provide for periodic:

replacement of the valves. This item will also include the implementation of system walkdown inspection requirements including specified frequency of walkdowns and documentation/reporting of walkdown results. -',,

CAPR 2 Description of corrective action to be A8930TT 03/01/01 tak~e~n, .:- (design Replace hem current design vacuum - approval/doc) breaker assembly with ase-protected configuration, Valve and Primer A8922MM Company Cat. ID 1036974 (6 inch), (physical work) 1036975 (8 iich)-1036976 (10 inch).

This CAPR should be accomplished coincident with CA i to minimize the number of valve replacements.

14

Actions Corrective Action Assignee Due Date CA1 Description of corrective action to be A8922MM 03/01101 taken:

Replace all vacuum breaker valve assemblies ii the CW Blowdown and Makeup Systems to restore system material condition. This CA should be accomplished coincident with CAPR 2 to minimize the number of valve replacements.

CA2 Description of corrective action to be A89301T 02101/01 taken:

Develop an adequate preventive maintenance ptograxn for the CW

.Blowdown' and Makeup Sys'tenm vacuum breaker valveaassemblies which includes periodic inspection of the valves including valve internals orpro'vides for valve replacement at appropriate intervals. This item will als 'ijclude the development of system' walkdown inspection requirements including specified frequency' ofwalkdowns and docimentatioporting of walkdown results. This program will be implemented by CAPR 1.

CA3 Description of corrective action to be A89100P 01105/01 taken:

Revise BwOP CW-12 to always require slowly op'ening thie' blow'down isolation valves (lt2CW018) in stags ove a few minutes when initiating blowdown flow from a no blQd'own;fhiw odi tion.

This CA does not apply whenfshifting blowdown from on unit to another as long as blowdown is not secured. This procedure revision shell beidentified as a corrective action per this'AT item.

15

-~ - . . . . . .. . . - . >, ,

^d. -

Actions Corrective Action Assignee Due Date

-CA4 Description of corrective action to be A8930TZ 05/01/01 taken:

Perform a review of a representative sample of preventive maintenance templates to determine if there are additional inappropriately assigned preventive maintenance templates.

EFRI Description of corrective action to be A8930TT 08/01/01 taken:

Perform an effectiveness review of the CAPRs.

F. Extent of Condition Other Exelon/Amergen Nuclear sites were'contacted to determine how those plants are configured for circ water blowdown and makeup and if they have experienced-any similar problems with vacuum breaker float assembly failures. -Byron and L.Salle stations were the only stations confirmed to have circ water blowdown and makeup systems that utilize vacuum breakers in their design. For CW blowdown and makeup systems, the extent of condition is limited to Byron and LaSalle.

Byron Station replaced their fiberglass blowdown and makeup piping in 1987 with carbon steel due to amline f associated with ground shifting. 12" Gol'den Aderson's model GH-7K vacuum breakers with line surge protection were installed at that time. System Engineering wilkdowns are conducted annually and Oprat relarly ives down th lies when they make River Screen House rounds. No vacuum breaker failusres have been identified and only minor amounts of water have been discovered contained within the valve vaults.

LaSalle's Operating Department performs inspections on their circ water blowdown and makeup systems on a semi annual basis. The inspections consist Iofleak hes and flushing the air releasevalve of any debris. The vacuum relief float is also checked and cleaned. Te ajoity f the problems they experience are related to plugging and freezing. The smaller air release valve float is the component that has frozenand it onlyaf.fcted the air'release aid rnot'th vacuum relief.- Corrosion has also been identified as an issue on the piping from the vacuum break to the air release valve. There is no history of vacuum breaker float assembly failures at LaSalle.

. . -h . -s -.-: vacuu

'. X H at;.- - -0. -.T:ae' - -

There are differences in system operation among the sites. In general, LaSalle operates their system continuously, whereas Byron and Bfaidwo od mhuistWcyclethe iblow'own systems to accommodate chemical feed additions.- Additionally, Byron still controls blowdown flow using te OCW 18A/B spray valves at their river screen house, thus their blowdown system is' maintained full at all times. Byron's spray valves are typicaly throttled to 20-22% open to maintain 12-13k gpm blowdown flow. The Byron spray valves are reportedly difficult to control and require frequent maintenance to remain operable.

16 n7Xll~~~~~~~~

nllccl oMlaIIu numen CI n7lXr

Other potential off site release patlis were evaluated for extent of condition. The summary of the evaluation is as follows:

NPDES Outfalls 001(a) - Wastewater Treatment Plant This system discharges to the U-2 Circ Water return line (to the lake)

-Discharge of waste' water from the non-radiological portion of the plant. This discharge consists mainly of:

Turbine Building Drains (TETF), Fire and Oil sump, Tendon Tunnel sumps, Aux Blr blowdown, Secondary side drains, Pretreatment drains and can be alternate discharge path for regenerant waste from Muds and CPs.

001(d) - Demineralizer Regenerant Waste

-Discharge of regenerant waste. This discharge consists mainly of:

Muds and C? regenerant waste, Regenerant chemical area drains (acid & caustic) and portable demineralizer regenerant waste (RO waste)

This discharges to the cooling lake via the circ water return to the lake 001(e) -River Intake screen backwash

-river screen house backwash is directed io the make-up pump suction (forebay) only during pump operation and is ultimately pumped to the lake with the make-up. This is includes as a point source only with no monitoring required.

002 - North Site Stormwater Runoff This discharge consists mainly of stormwater from the Parking lots, Transformer areas, Station aid North Site area as well as Station roof drains. The discharge path is 'ia the runoff ditch along the west side of the property and is ultimately discharged to the Mazon River. No monitoring is required for stormwater 003 - South Site Stormwater Runoff This discharge consists of stormrwater runoff from the area south and east of the main site (near the LSH). The discharge path is via the runoff ditch along the west side of the property and is ultimately discharged to t1he Mazon River. No monitoring is required for stormwater 004 - Switchyard Area Runoff This discharge consists of storm'water runoff from the switchyard area. An oil separator exists in this area to remove oil in the event of leakage from the switchyard equipment.

The discharge path is via the runoff ditch along th weste side of the property and is ultimately discharged to the Mazon River. No monitori g is required for stormwater 17 a- -n . n.,aS U ffFI ... I -. Ibu- - 0.-. c,,..

There are two special conditions that identify flowpaths not included as 'normal' discharges.

Special Condition 9: Discharge of station cooling pond water to adjacent impoundments owned by the permittee, to replace water which is withdrawn from these impoundments for station operation during periods of low flow in the Kankakee River when the station must decouple its operation from the river, is hereby permitted for these emergency periods.

No monitoring is required however the Agency (IEPA) must be notified if this occurs.

This simply says we can refill Monster Lake if we have to pump it down during drought conditions.

Special Condition 12: An emergency cooling pond overflow exists tributary to an unnamed drainage' ditch that is tributary to the Mazon River. Discharges to this overflow shall be subject to the bypass provisions of 40 CFR 122.41(m)

-This states that monitoring of 001 (Cooling Pond Blowdown) parameters must be performed daily during discharge.

Additionally the Make-up to the lake (river water) would be a potential source of 'discharge' should we develop a leak in that line.

The NPDES related treatment systems discharge either directly into the blowdown line or to the Circ Water return to the lake and then ultimately discharge through the blowdown.

The Wastewater Treatment and Sewage Treatment processes are separated from the Main Plant or Turb/Aux buildings. Any leakage or failures within these systems would be contained within their respective buildings and/or general area. The Demineralizer Waste process is contiained within the Turbine Bldg and the Radwaste systems aIe located prirnarily'within the Aux B however some components (release tanks) are located in the Turbine building. Any leakage or failures within these systems would be contained witiin their respective areas.

The connections to the Circ Water return to the lake (Wastewater and Demin Waste) are located undeigr6ufnd in the area west of the Turb Bldg. The Sewage Treatment Plant connection is:

underground in the area north of the"Sewage Plant'and the radwaste tank discharge ties in to the Ities

, , late.aige blowdown within the Turb/Aux Bldg. Any catastrophic failure of these cornectiofs should in to thte be visible by localiz bubbling or sauftition of the grouid in which case the discharge could be isolated. This would prompt investigation by excavation to determine cause of failure and initiation of repairs. Preventive measures could be taken at that time (berms or other barriers) to prevent this leakage from entering the runoff ditches.

Periodic rounds of these systems should provide the early detection of any unusual conditions.

Preventive measures could then be taken' to eliminate potential for release of water frfom these systems.

18

%7 jig I nace9e e0 COOMDIt YIJU I auUrnb

G. Risk Assessment There were no plant specific risks associated with this issue. There were no risks to the CW Blowdown system as a result of this issue, since in the failed condition the vacuum breaker assembly would still function to prevent a vacuum from forming and causing damage to the blowdown piping.

PREVIOUS EVENTS:

Search of CAP, EWCS, OPEX, etc. was conducted. The OPEX search was limited to the previous 2 years and confined to the variables of "unplanned releases of liquids", "vacuum breaker failure and releases", and "blowdown vacuum breaker". The EWCS search was limited to both Circ Water Makeup and Blowdown vacuum breaker failures. The CAP system search was limited to the Circ Water System. No similar vacuum breaker float failure events were identified via these searches.

EVALUATOR COMMENTS:

The station was slow to implement Event Response Guidelines, CWPI-NSP-AP-l-l, or NGG Issues Management, OP-AA-191-503. The initial facts of water in the ditch on Smiley Road and the source of the water being a leaking vacuum breaker on CW blowdown piping were known at approximately 1600 on 11/06100. NGG Issues Management was not entered until soretime on 1l/09/00 after results of radioactive sampling showed both isotopic and Tritium samples above LLD, a delay of about 3 days. The decision to form a root cause' team to investigate the issue was not made until 3 days after the event. The station did establish separate teams for event recovery and root cause.

Interviews with station management personnel responding to the initial event reports revealed there was no consideration for use of the Event Response Guidelines or NGG Issues Management procedures for general guidance. OP-AA-101-503, NGG Issues Management was developed as a corrective action to the Braidwood Oil Spill Event in order.to respond appropriately to significant issues, including those that have potential media or public interest.

Since it was known that the blowdown line'had most likely been leakinfor more than a week (based on the information from the resident who reported the water inmthe' Smiley Road ditch) and that radioactive releases from the Site are conducted via the blowdown'line. the potential for radioactivity in the Smiley Road ditch should have prompted entry into the NGG Issues Management Procedure due to the possibility of public and media interest.

Entry into the NGG Issues and Management Procedure did eventually occur on 11-9-00, three days following the initial event. The' station did determine that a prompt investigation was not required.

The establishment of separate teams for event recovery and root cause is an apparent lesson learned from the Oil Separator event (CR # A2000-2683) which occurred in June 2000.

Condition Report #A2000-04465 has been written to document slow station response in implementing NGG Issues Management or Event Response Guidelines.

19

" rXen 1-v/Vii Ad OWoz L0[C2C*CIII poompJ3Jla ji11)4 flNMrn

l l t-b

• s t

Braidwood Nuclear Station Discharge Pipe Vacuum Breaker Leak Dose Assessm-ent

'I.-.

Performed By:

Richard W. Dubiel, CIP JosephDarFMan, CHP Millenniuxn Servces, hic.

.222 Creekstone Ridge Woostbck, GA 30188 Submitted By:

Ricrd W. Dubiel February 19,-2001 Rev. 1

Braidwood Nuclear Station Discharge Pipe Vacuum Breaker Leak Dose Assessment

Introduction:

In the fall of 2000, a vacuum breaker on the mai discharge pipe was found to be leaking.

The discharge pipe directs a high volume of non-radioactive effluent as a result of normal station operations. The discharge line also serves as a dilution volume for normal station radioactive discharges These dischaiges ar made in accordance wiih the station technical specifications following samlin a sis, and confirmation that adequate and dilution fow is available Following eyof vacuum breaiS leak, soil saiples were obtained in the vicinity of tie leak. Thesamles were obtained and analyzed by Braidwood Station personneL E samling program was des to detemine the ex t of s soil that mnay hive be n cna ted with low leve of Active niaterai The vacuuineMakeris oused in a concrete vaumTe hotto noftthe vault is not led. Water was observed ruinrng from the Manway cover and onto the suhce ofthe ground. A spling grid was established minthe viniofthe valt.- Soil samples were obtained from the top 6 inches of soil at the approximate center poi ofeach gd. No core bores were obtained in the samig grids -or vithinite vault.- Tie prnarypurpose of his assessment is to evaluate the potentialydose due to the repoti radioiit fno e diaactions are taken Additiol report ev es cmpiance ith certain regulations, aniday beused6to document s in accordance with I OCFR50.75(g).

Methodology:

Soil aples were obtained and analyzed by Braidwood Nuclear Station personnel A sampling grid consisting of 10 meter by 10 meter rids was established around the vault.

Soil aples of the top 6 inches were obtained fro the approximate center point of the grds. Samples were taken in"all directions, diayoutwad until an s resilts showed less thaminimum detectable activity (MDA). The maximum MDA for th soil

.5E-07 e for iotoes conideed. This MDA ivle willresult in a loiverMDA in morete~isilyidextified isotopes (e.g. Co-60). Figure 1 provides a simplified layout ofthe samplig gnrd. Grids with white backgrounds indicate those grids

-inmg-sampleanalysis results ndica at least one isotope greater than MDA. Tose grids wAi dark bagronds ndicate ihose grids with nio resuls greater than MDA.

Table I -provides the isotopes activity level for each a zed isotope i reported reslts greater than MDA4 by grid location.

The quantity of each reported radionuclide was estimated to determine the need for posting the area in accoIdane Rio O.1902(e). Eah sie rest was assumed to be rpresentative ofthe grid area. For saples withpositive results, the conceation of each reported isotope was assumed to bniform distributed over tie 100 m2grid areas to a depth of 15 cm. Table 2 provides the calculations and the suimfation of activity in the area.

2

A dose assessment was performed using RESRAD. 6.0, created August 25, 2000. The RESRAD Version 6.0 computer code was developed under the joint sponsorship of the U.S. Nuclear Regulatory Conmmission and the U.S. Department of Energy for site-specific dose assessment of residual radioactivi The computer code was developed at the Environmental Assessment Division of Argonne National Laboratory. The'RESRAD list of isotopes includes both Co-60 iand Mn-54, but does not contain either Co-58 or Te-123m. Co-58 and Te-123m were assessed using fitors from other codes as described below.

RESRAD provides defiault parameters for dose calculations, but allows fbr site specific pammeters for a specific situation to be applied. RESRAD will evaluate the dose associated with all paways icudg direct radiation, water pahway, food y etc. For the specific isotopes identified in the soilales, the direct radiatiofn iathway is dominant, accounting for more thah 98% of the dose in the critical frs't year. Therefore for all pathways other than the direct radiatonWath` default parameters were used, since little impact on the dose would result fm sitespecific parameters.

To evaluate the dose associated with the direct radiation from the radioactivity i the soil, the citisca paeters are the average activity of chin the ted zone, the surface area (squarMeter areaco gthe material), the depthofthe contamination, and depth of clean cover material if any. No clean cover matertal was is e. TIh depth of the contaminated zone was set at 15 cm. To provide a*. 0 conservative estimate of the actfliy, the average radioactivity of each isotope identified was calculated, using only those sales ip ositive values. For example, the vlue for Co-60 was the average of the four sap positive values for C-60 M-54 two positive values were used, etc. Although not ihkluding the samples with vlues less thin MDA will introduce a conservative bias, this approach is not co sidrd y conservative. The bias is not considered to be substantil snce the eal dihstion of contami on, with minmum grid size of 10 meters by 10 meters (100 n), ;does not-haVe a significat impact on the dose. cted surc area was consideeto be -60 mes eters (3600 n2), an area that encompasses h full extent of grids with sample analysis results indicating greater ta MDA. gain, assuinig this area introduces a conservative bias. The bias introduce byi assuming 3600 m2 rather than a iniiimum area of 100 m2 is approxhnatel20%.%.

Calculations were made using the parameters described above to determine a co rve estimate of dose. Calculations were also madeuiing the imum concentrations for each isotope to detcrnine a bounding value to show compliance. The ress are presented i the "Results" section below. Te RESRAD cculations provide the annualized dose due to Co-60 and S-54. Co-58 and Te-123n are not included in RESRAD suite ofnuclides, ap ue to tir relatvely shor half-lives. To determine the dose due to these isotopes comparisons were made to Co-60 using the NRC default values fromNUREG/CR-5512 and the NRC computer codefor screening values in soil, DaindD2. The ratios derived from each of these hods and ihe relatie dose contribution from Co-58 and Te-i23m are presented in Table 3.

3

Results:

The output of the RESRAD computer code dose calculations are provided as Appendix I and are summarizid in Table 4. Table 4 presents the dose from each isotope for tlhe average activity and for the bouiding condition using the maximum activity each nuclide. The best estimate, based on the average lues of Co-60 and measured for Mn-54 rehilt in a maxrm dose in the rst year ofl1.225m rem. Additional dose in the first year, due to the short lived nuclides, based on'their average concentration and dose' fctor as related to Co-60 are: Co-58, 0.121 mrem, and Te-123m, 0.030 mrem. The total dose during the critical first year is 1.376 mrem 98.6% of that dose' is due to direct radiation from the radioact tivity in the i ere sice. Due tthe short half lies of all isotopes idetified, the dose'in subsequient years is a reaction ofthe and is due almost entirely to Co-60. The dose due to' Co-60 adi Mn-54.Jduiiii the second 0.995 me Contribtion from Co-S8 aid Te-123m are less than .1% during year is the second year due to their short halef ves. Dose durig squnt years il decrease onsistent with the half-life of Co-60. The subsequent year dose values areprovided page 8 of the'RESRAD outtreport. No'te tat t dose value desig""atedmeanse in Appendix A, integrated dose in the year foowing thie year designated. Figiure'2 provides do's'e':due 'to' 'Co-60 and -54 as a futio o'ftime. Note tht Figiure a graph'ofthe 2 is the' standard RESRAD grahic outputaind does not nclude the iiial year dose estimate. The dose due to Co-58 and Te-l23m following the'fi year s less a 0.1 mrem for each istope, and essentially zero following tlie second year.

The doses due to pathways otheri than direct exposure, i.e. water and food pathways, etc.

is less 1.4% of the total dose etiae, sin anmaximum dose of less tbai 0.019

'mrem in 'te first year.' Exosure due to e activi the soil Wi b red individuils within the owner-contoled area. Onyrtions of the non-direct intly to pathways are capable of resulting in exposure to individuals beyond exposure hde' site bound Therefore, there isno potential for violation of1ocFRo. Appenix I entire non-direct exposure patw limiiits, even if the e was assu Ato be beyond the site'boui*r, and occurring within a single quarter.

The bo ig dose calculation assumes that all of the 3600 zi2 area'6cntains at the aimm concentrationimeasured for eachisotope. -This calculation radioactivity is a con proach' used for conp ce evaluation,'and is not considered to best imate ofthe dose. The uingv be the eusingRAD Versioi 6.0, 'uses te same parameters other than isotope concentrationas used inthe parevious and ratios for Co-58 and Te-123n as previoity esablished, yields a bo al tions, dose value of 2.21 nmrem inthe first ye ar.

An estimate of the totil activity was' also pef ormed for comparisonwi the atities requiring posting under 10CFR2O.1902(e). The activity of each isotope was deteined bynijiltiplying the co iaion of that isotope easu in each grid, in'iCi per gram of soil, Vthe mass, in gr#amsofthe soil in the grie soil was6assumed to- b unifo contamated to a depth 'of 15 c. 'Ech gIid is i0 meters by 0 meters.

The density of soil is assumed to be 1.6 gmi/cc, consistent with the deliit'vlue ofRESRAD.

4

The total activity of each isotope and a comparison with the limits specified in I OCFR20 are provided in Table 2. The Co-60 activity exceeds the level requiring posting as a Radioactive Materials Area Summary:

The assessment of the effluent discharge line vacuum breaker leak resulted in the following:

j!306

• The dose assessment for the reported radioactivity in the soil pounding the vacuum breaker access vault results in a maximum dose of 404 nremn in the first year. 98.6% ofthe dose is due to direct radiation exposure form the gamma radiation.

• The relatively short half-lives of the nuclides present results in a dose reduction in subsequent years. Following the second year, the dose reduction is consistent with the half-life of Co-60, 5.26 years.

• The results of this assessment indicate that the non-direct exposure pathway doses are less than the limits specified in IOCRF5O, Alppend I.

• The results of this assessment indicate that the direct exposure rate in the vicinity of the vacuum breaker access Viiut is in compliance wffi IOCFR20.1301 and 20.1302.

• The quantity ofradioactive material in the soil surrounding the vacuum breaker access vault exceeds 10 times the values specified i Appendi C to 10 CFR20.

The area requires posting in accordance ith 10CFR20.1902(e).

• The licensee should evaluate the requirements of Subpart I, Storage and Control of Licensed Material 5

- I I i t *fBM-h*I

Figure 1 SAMPLING GRID NORTH

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. . . P f :' 0 -- ,9

. . -: Ah'

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. ,,, '.J

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-,.- ., -,, , 's .- -i i, .

GridSiuplJ 6

Table 1 Soil Samplinf'Isotopic Results Grid isotope . Results (pCi/g) 3 Co-58 '2.5E-07 Co-60 2.3E-07 Te-123m 1.2E-07 4 Co-58 9.8E-08 Te-123m 1.2E-07 6 Mn-54 9.9E-O8 Co-58 3.3E-07 Co-60 I.9E-07 Tc-123m 3.7E-07 12 Co-58 7.6E-O8 13 Co-60: 6.OE-08 -

1 Te-123m 2.8E-08 16 Co-58 5.3E-08 18 Co-58 .OE-07 Te-123m 5.5E-08 20 . __8-4.9E1-08 30 O-58 5.OE-08 35 Mn-54 1.4E-07 Co-58 8.9E-07 Co-60 2.2E-07 Te-123m 8.1E-07 62 Co-58 l.OE-07 63 Co-58 9.5E-08 7

Table 2 Total Activity Calculation Total Total Total Total Isotope Mn-54 Mn-54 Co-58 Co-58 Co-60 Co-60 Te-123m Te-123m Grld# pCi'g gci pCvg Fci 'pCyg FCi pCi/g giC 3 0.25 6.00 0.23 5.52 0.12 2.88 4 0.098 2.35 0.12 2.88 6 0.099 2.38 0.33 7.92 0.19 4.56 0.37 8.88 12 0.076 1.82 13 0.06 1.44 15 0.028 0.67 16 0.053 1.27 18 0.1 2.40 0.055 1.32 20 0.049 1.18 30 0.05 1.20 35 0.14 3.36 0.89 21.36 0.22 5.28 0.81 19.44 62 0.1 2.40 63 0.095 2.28 Total :I 5.74 50.18 N 16.8 36.07 Limit 1000 1000 10 100 Soil density = 1.6 g/cc Mn-54!App. C = 100. uCi Co-58 App. C - 100 uCi.

Co-60 App. C - 1 uCi TeA123m App. C = 10 uCi 8

Table 3 Relative Dose Contribution from Co-58 and' Te"-123m Isotope NUREG/CR-5512 Ratio to Co-60 DandD2 Ratio to Co-60 Screening Values Nob (1) Normalized Dose - Note (2)

(pC!/gin) (mrem/yrlpCi/gm)

Co-60 3.79 1 6.6 Co-58 34.7 0.109 0.72 0.109 Te-123m 185 0.0205 .135 0.0205 Note 1: Co-60 screening value divided by individual isotope screening value from NUREG/CR-5512 Note 2: Individual isotope calculated dose divided by dose due to Co-60.

Calclations perfonned using DandD2, 1 pCi/gn, 1year perod, defult parameters.

9

.;Table'4 Dose Assessment Isotope Half-Life Average Conc. ose - Maimum Bounding Dose Assessment Conc. Assessment

._._. (Note 1). (Note 2)

Co-60 5.26 years 0.18 pCi/gm 1.056 mrem/yr 0.23 pCi/gm 1.350 nirem/yr Mn-54 303 days 0.12 pCi/gm. 0.169 mremWyr 0.14 pCi/gm 0.197 rnrem/yr CM-58 71.3 days 0.19 pCi/gm 0.121 mrem/yr 0.89 pCi/gm 0.569 mren/yr Te-123m 117 days 0.25 pCi/gmi 0.030.mremn/yr 0.81 pCi/gm 0.097 mrem/yr Total - 1.376 mremlyr Total 2.21. mrem/yr Note 1: All doses are for year. 1, maximur dose year. Co-60 and Mn-54' dose from RESRAD Version 6.0. Factors applied for CO-58 and Te-123m as-ratio to6Ca-60 dose:from Table 3.

Note 2: Bouhdingvalues are based on'maximum measured concentration for each isotope applied over entire area (3600 n 2 ).

10

-Figure 2 Time Dependent Dose Due to Co-60 and Mn-54 DOSE: All .-Nuclides 'Sum'med, All Pathways Summed L0.5 E

00

. 1 10 100 1000 Years I Co-60' e Mn-54 -8 Total BRAIDI RAD 02/20/2001 14:03 Includes All Pathways 11

Appendix 1 RESRAD -6.0 RCalculation Output . .Report V

tc/tt InlrectcIat to om l goJ1 II VI 11ijklon

ESRAD, Version 6.0 T< Limit = 0.5 year 02/20/2001 14:03 Page 1 ammary : Braidwood with NRC Recomended Default Parameters File: BRAID1.RAD Table of Contents Part I: Mixture Sums and[Sinnle Radionuclide-Guidelin s fttfI~ff~tfftffffftfffIfiIf ift+/-tffftffIfffftItfffIffrft Dse Conversion Factor (and Related) Parameter Summary... 2 Lte-Specific Parameter Summary ........................ 3 runary of Pathway Selections .......................... 7

)ntaminated. Zone and Total Dose Summary .................. 8

)tal Dose Components Time;> O.OOOE+00 .. ................................... 9.

Time 1.000E+00 ...... 10 Time 3.OOOE+00. 11 Time 1.000E+01: .................................... 12 Time 3.OOOE+O1 ................................... 13 Time 1.000E+02 .................. 14 Time - 3.000E+02 .................. 15 Time = 1.000E+03 .................................... 16 ise/Source Ratios Summed Over All Pathways .............. 17

.ngle Radionuclide Soil Guidelines ........... 17 iae Per Nuclide Summed Over All Pathways .... 18 ail Concentration Per'Nuclide ................... 18 I

I I .

Appendix 1 Page 1 is

kESRAD, Version 6.0 Tec Limit = 0.5 vear 02/20/2001 14:03 Pacqe 2 lummary : Braidwood with NRC Recomended Default Parameters File: BRAID1.RAD Dose Conversion Eactor (and Related) Parameter Summary File: Default.LIB

3. Current aue~ ' Defaul 3 Parameter Name leu armte 3 1-1 3 Dose conversion factors for inhalation, mrem/pCi: .

-l 3Co-60 ' 2. 190E-04 '-2.190B-04 ' DCF2( 1)

.-1 ' n-54 ' 6.700E-06 J:6.700E-06 .DCF2( 2) 3 3 3.

• -1 ' Dose conversion factors for inqestion. mrem/pCi: 3 '

• -I 3 Co-60 '..2.690E-05 ' 2.690E-05 3 DCF3( 1)

• -1 ' Mn-54 S 2.770E-06 '2.770E-06 3DCF3( 2) 33

-34 3 Food transfer factors: 3 3

-34 3 Co-60 . plant/soil concentration ratio, dimensionless 3 8.000E-02:' BA.OOE-02 ' RTF(- 1,1)

-34' Co-60 , beef/livestock-intake ratio, (PCi/kq)/(pCild) ':2.OOOE-02:' 2.000E-02 3'RTF 1-,2)

-34 Co-60 , milk/livestock-intake 7atio, (DCi/L)/(pCi/d) 3 2.000E-03 3 2.00OE-03 3 RTF( 1,3)

-34 ' , ,

-34 3 Mn-54 . plant/soil concentration ratio, dimensionless ' 3.OOOE-O1 3 3.000Z-01 1 RTF( 2,1)

-34 'Mn-54 , beef/livestock-intake ratio, (uCi/kcT)/(nCi/d) ' 5.OOOE-04 3 5.000E-04 3 RTF( 2,2)

-34 'Mn-54 , milk/livestock-intake ratio, (uCi/L)/(pCi/d)' 3.300OE-04 33.000E-04 3 RTF( 2,3)

-5 'Bioaccumulation factors, fresh water. L/ko: 3 '

-5 3Co-60 , fish 3 3.000E+02 3 3.000E+02 ' BIOFACC 1,1)

-5 ' Co-60 , crustacea and mollusks 3 2.OOOE+02 3 2.000E+02 3 BIOFAC{ 1,2)

-5 3 3 3 3

-5 'Mn-54 , fish '3 4.OOOE+02 3 4.OOOE+02 3 BIOFAC( 2,1l

-5 bMn-54 , crustacea and moll.usks 39.OOOE+04 39 OOOE+04 2s BIOFACJ.Z,2)

Appendix 1 Page . 18

ESRAD, Version 6.0 Tac Limit = 0.5 vear. 02/20/2001 14:03 Paqe 3

• ummary Braidwood with NRC Recomended Default Parameters File: BRAID1.RAD Site-Specific Parameter Summary

. 'User ' , Used bv RESRAD 3 Parameter enu ' Param'eter Default, 3-*not~ 3 (If different~from user irnput) 3 Name 011 ' Area of contaminated zone (m**2) 33.600E+03 S 1.OOOE+04

3

'AREA 011 T Thickness -of contaminated zone (m) .'500E-012.000E+

. ' --- ' THICKO 011 ' Lenqth parallel,'to aquifer flow:.m) ' 1.0OOOE+02.':1t'000E+02:3 LCZPA02 01 1 ' Basic radiation dose limit Crrem/yr) , 2.500E+O.'. 2..500E+O1 ' , --- - BRD 011 ' Time since placement'of material (vr) '. 0.000E+00' O.OOOE+00 ' --- ' TI 011 ' Times for calculations (vr) ' l.OOOE+OO s l.OOOE+00 3 3 T( 2) 011 ' Times for calculations (vr) 2 3.000E+00 ' 3,000E+'00 -- Tf 3) 011 ' Times for calculations .(yr) ' 1.OOOE+01' 1.OOOE+01 ' - T( 4) 011 ' Times for calculations (vr) 3 3.000E+01 ' 3.000E+01 3 - T( 5) 111 ' Times for calculations (vr) .3 l.'OE+/-02. ' 1.OOOE+02 ' ' T( 6) 311 ' Times for calculations (vr) ' 3.000E+02'3 3.000E+02 ' ' T( 7) 311 ' Times for calculations (vr) 2 l.,000E+03 ' 1'.000E+03 ' --- 3 T( 8) 311 ' Times for. calculations (vr) ' not used 0OOOE+ 00 . --- 3 TC 9) 11 3 Times for calculations (vr) 3 not used I 0.000E+00 3 --- 3 T(10) 3 3 . . . a 312 3 Initial principal radionuclide.(pCi/a): Co-60 3 1.800E-01 ' 0.OOOE+00 3 --- a Si( 1)

)12 ' Initial principal radionuclide fpCi/a): Mn-54 ' 1.200E-01 ' 0.000E+00 - . ' 51! 2)

)12 3 Concentration in qroundwater (vCi/L): Co-60. 3 not used ' 0.OOOE+00 ' ' WI! 1)

)12 3 Concentration in aroundwater (pCi/L): Mn-54 ' not used 0.0.00E+00 3 ' W1( 2) a 3 . 3 3

3

)13 3 Cover depth (m) 3 0.OOOE+00 O.OOOE+OO

)13 ' Densitv of.cover material (q/cm**3)

' - COVERO

' notused ' 1.500E+00 3 --- ' DENSCV 113 ' Cover depth erosion rate (m/vr) ' not used  :' 1.000E-03 ' . VCV

)13 ' Densitv of contaminated zone (q/cm**3) ' 1.500E+00 ' 1.500E+00 ' --- 3 DENSCZ 313 ' Contaminated zone erosion rate (m/vr) 3 1.OOOE-03 1.000E-03 - 3 VCZ 113 ' Contaminated zone total porositV 3.4.000E-01:' 4.000E-01 3 --- ' TPCZ 113 ' Contaminated zone field capacitY ' 2.0 0 200'E-01 2.0 3 --- J FCCZ 113 ' Contaminated zone'hvdraulic conductivity (m/vr) ' l.000E+01 ' 1.000E+01 ' --- . HCCZ 113,' Contaminated zone b Parameter ' 5.300E+00 5.300E+00 ' -' BCZ 113 ' Averaqe annual wind speed (m/sec) 3 2.000E+I00 ' 2.0QOE+00 3 --- ' WIND 113 3 Humiditv in air (a/m**3) 3 not used I 8.000E+00 - HUMID C

'13 ' Evapotranspiration coefficient . ' 5.000E-01 ' 5.000-01 ' ' EVAPTR

'13 ' Precipitation (m/vr) 9 1.000EfO '.OOOE+00 ' --- 3 PRECIP C 13 ' Irrigation (m/vr) ' 5.000E-01 ' 2;000E-01 ' - --- ' RI

13 ' Irrioation mode ' overhead overhead --- 3 IDITCH 3

13 ' Runoff coefficient a 2.000E-01 ' 2.OOO-101 -- 3 RUNOFF 13 3 Watershed area for nearbv'stream or pond (m**2) I 1.OOOE+06 1 1.000+E06 ' --- ' WAREA 13 ' Accuracv for water/soil computations ' l.000-03 '1.000-03 ' --- ' EPS b 14 3 Densitv of saturated zone (q/cm**3) 3 1.500E+00 3 1.500E+00 a --- a DENSAQ 14 3 Saturated zone total vorositv ' 4.000E-01 3 4.OOOE-01 3 --- X TPSZ.

14 ' Saturated zone effective porosity 2 2.000E-01 ' 2.000-01 ' -- ^ ' EPSZ 14 3 Saturated zone field capacity a 2.000E-01 3 2.00OE-01 3 --- ' FCSZ 14 ' Saturated zone hydraulic conductivity Im/yr) ' 1.000E+02-.' 1.OOOE+02 ' --- ' HCSZ c 14 3 Saturated zone hydraulic gradient ' 2.000E-02 ' 2.000E-02 ' --- ' HGWT 14 3 Saturated zone b parameter 3 5.300E-00 ' 5.30000 ' -- ' BSZ 14 3 Water table dron rate (m/vr) 3 1.000E-03 1.OOOE-03 ' ' VWT 14 ' Well pump intake depth (m below water table) ' 1.000E+01. ';1OOOE+01 ' --- 3 DWIBWT c 14 '-Model: Nondispersion (ND) or Mass-Balance (MB) ' ND ' ' 'ND ' 3 --- 3 MODEL 14 ' Well pumping rate (m**3/vrl .' 1.180E+02 3 2.500E+02 ' a UW 3 3 3 3 3 Appendix 1 Page 18

ESRAD, Version 6.0 T<< Limit - 0.5 vear 02/20/2001: 14:03 Paqe 4 ummary Braidwood with NRC Recomended Default Parameters File: BRAIDI.RAD Site-Specific Parameter Summary (continued) 3 User 3 3 Used bv.RESRAD . Parameter enu ' . Parameter  : Input Default 3 different fromuserinput)':

3 'tIf *Nam 015 2 Number of unsaturated zone strata' 1. 34 '

015 '3 Unsat. zone 1, thickness (m) ' 4.OOOE+00, 3 :4,000E+00. ' ' Hl(l) 015 -Unsat. zone 1, soil densitv (q/cm**3) '.500E+00 3 I1.500E+00:3 - DENSUZW1) 015 3 Unsat. zone 1, total porositv ' 4.000E-01 ',4.000E-O1 ' . ' TPUZ(1) 015 3 Unsat. zone 1. effective porositv '2 OOOE-01,'2.0'000E-1 ' . EPUZ(1) 015 ' Unsat, zone 1, field'capacitv 3.'2.000E-01:'2.'000E-01 --- ' FCUZC1) 015 ' Unsat. zone 1, soil-specific b parameter X 5.300E+00 ,5.300E+00 ' BUZ(1) 015 2 Unsat. zone 1, hvdraulic conductivity (m/vr) 3 1.OOOE+01 :1.000E+01 ' --- ' HCUZ(1) 3133 3

016 3 Distribution coefficients for Co-60 3

)16 ' Contaminated zone (cm**3/a) . ' 1.000E+03 ' 1.000E+03 3 -- DCNUCC( 1)

)16 ' Unsaturated zone 1 (cm**3/q) 3 1.000E+03 ' 1.OOOE+03 ' --- I DCNUCUl 1,1

)16 3 Saturated zone (cm**3/p) 1'.OOOE+03: 1.000E+03 '.DCNUCSC 1)

)16 ' 'Leach rate (/vr) ' 0.000E+00 ' 0.000E+00 3 2. 8BE-03 3 ALEACHf 1)

)16 3 .Solubilitv constant ' 0.OOOE+00 3 0.000E+00 3 not used 3 SOLUBK( 1) 3 3 3 , 3 ' 3

)16 'Distribution coefficients for Mn-54 3 ' ' .

)16 3 Contaminated zone (cm**3/a) 3 2.00OE+02 '3 2.000S+02 ' --- 3 DCNUCC( 2)

)16. Unsaturated zone 1 (cm**3/a) 2.,000E+02 2.OOOE+02. - DCNUCU(

)16 ' Saturated zone Ccm**3/cz) 3 2,1 I 2.000E+02 ' 2.000E+02 --- 3 DCNUCSC 2)

)16 3 Leach rate (/vr) 3 O.OOOE+00 0.0005-1+00 ' 1.443E-02 3 ALEACH( 2) 116 33 Solubilitv constant 3 0.000+-00 '0.000E+00 2 not used ' SOLUBC( 2)

S a . 3 117 3 Inhalation rate (m**3/vri ' 1.169E+04 3 8.400E+03 ' --- 3 INHALR 117 3Mass loadina for inhalation (q/m**3) ' 3,140E-06 ' 1.OOOE-04 ' - MLINH' J17 ' Exposure duration 3 3.000E+01-: '3.000E+01 ' - ED 117 ' Shieldinq,factor, inhalation .3 4.000E-01 ' 4.000E-O1 ' - SHF3 117 3 Shieldina factor, external qamma ' 5.512E-01 ' 73.000E-01 ' - SHF1 117 ' Fraction of time spent indoors .' 6.571E-01.' 5.000E-013 FIND i17 ' Fraction of time sDent outdoors (on site) l.101E-01 2.500E-01 ' - 3 FOTD 117 3 Shape factor flaQ, external qamma ' 1.000E5100 1.0005+00 >0 shows circular AREA., FS 117 ' Radii of shape factor arrav (used if FS -1): 3 '

117 ' Outer annular radius Wm). rina 1: I not used ' 5.000E+01 3 - RAD SHAPE(

'17 3 Outer annular radius Cm), rinQ 2: ' not used ' 7.071E+01 ' --- ' RAD.SHAPE(

'17 ' Outer annular radius Cm). rinq 3: 2 not used I 0.000E+00 ' - ' RAD SHAPE(

'17 3 Outer annular radius (m), rinq 4: ' not used 2 0.000E+00 3 RADSHAPEC 17 3 Outer annular radius (m), rino 5: 3 not used . 0.000E+00 3, RAD SHAPE ' C 17 3 Outer annular radius (m). rinq 6: ' not used ' 0.000E+00 3 RAD SHAPE(

17 3 Outer annular radius lm). rinq 7; ' not used 3 O.OOOE+00 I 3 RAD SHAPEC 17 3 Outer annular radius tm), rinq 8; 3 not used I O.OOOE+00 3 --- ' RAD SHAPE(

17 3 Outer annular radius Cm), rino 9: ' not.used ' 0.000E+00 I -- RAD SHAPE(

17 ' Outer annular radius (m). rinq 10: ' not'used 3 0.000E+00 3 3 RAD SHAPEU1 17 3 Outer annular radius (m). rinq 11: 3 not used a0.000E+00 3 3 RAD SHAPE(l 17 3 Outer annular radius (m), rinq 12: ' not used '0.000E+00 ' --- 3 RAD SHAPE1l 3 3 3 3 3 Appendix I Page i 18

BSRAD, Version 6.0 Te Limit 0.5 year .02/20/2001 14:03 Pace 5 immary : Braidwood with NRC Recomended Default Parameters File: BRAIDI.RAD Site-Specific Parameter Summary (continued) 3 User 3 UTsed.bv RESRAD ' Parameter aamtr~Input- 3n *2 Default- 3' (If :different' fromn-user input)' -Name 33

)17. ' Fractions of annular areas within AREA: - 3 3 117 ' Ring 1 not used- '.l.OOOE'+00 - ' FRACA( 1)

)17 3 Rina 2 3 not used ' 2.732E-01 3 --- 3 FRACA( 2) 117 ' Ring 3 ' not used 3 .O.000E+O.' -- FRACA( 3)

)17 ' Rinq 4 ' not used '0.OOOE+OO ' --- . FRACA1 4) 117 3 Rina 5 a not-.used: a: 000E+00T3 3 FRACA( 5) 17 3 Ring 6 'not used. 3.0OOOE+00 ' --- FRACAC 6) 117 3 Ring 7 ' not used . O.OOOE+00 3 --- ' FRACA( 7) 117 3 Ring B 3 not.used. '0O.OOOE+00 3 --- 3 FRACA( 8) 117 3 Rina 9 3 not used ':0.000E+00 ' -- 3 FRACA( 9) l17 ' Ring 10 3 not used: .3 0.000E+00 . 3FRACA(lO) 17 3 Rinq 11 3 not used ' O.OOOE+00 . -- 3 FRACA(II) 17 Rina:12 ' not used ' 0.OOOE+00 ' --- ' FRACA(12) 3 3 3 3 I 118 Fruits, vecetables and arain consumption (kq/vr) 3 1.120E+02 '. 1..600E+02. . 3 DIST(l)

• 18;' Leafv veaetable consumption Ckq/vr) 3 2,140E+01 ' 1.400E+01 3 DIET(2) 18 3 Milk consumption (L/vr 'A2.330E+02 '.9.200E+01 3 DIET(3) 18 3 Meat and poultrv consumption (kq/vr) ' 6.510E+0l '6.300E+01.' ' --- DIET(4) 18 ' Fish consumption (ke/vr) 3 2.060E+01 ' 5.400E+00 'I --- DIET(5) 18 3 Other seafood consumption (kq/vr) ' 9.000E-Ol:: 29.00OE-Ol ' 3 --- DIET(S)

IS 3 Soil inqestion rate (a/vr) .' 1.626E+01 ' 3.650E+01 - 3 SOIL 18 3 Drinkina water intake (L/vr) . '.4.785E+02 3 5.100E+02 ' --- .' DWI 18 3 Contamination fraction.of drinkine water 3 1.000G+00 '11.0O0E+00 ' - FDW 18 3 Contamination fraction of household water 3 not used ' 1.OOOE+00 ' - ' FHHW 18 3 Contamination fraction of livestock water 1.000E+00 ' 1.000E++00 --- ' FLW 18 3 Contamination fraction of irriaation water ' 1.000E+00 ' 1.OOOE+00 --- 3 FIRW 18 3 Contamination fraction of aouatic food 3 1.000E+00 3 5.000E-01 3' --- FR9 l8 3 Contamination fraction of plant food 3 1.00+00 '-l 3 -- FPLANT 18 3 Contamination fraction of meat ' 1.000+002'-l. 3 -- FMEAT 18 '-Contamination fraction of milk ' 1.000E+00 '-1 3 --- 3 FMILK 3 3 . 3 t 19 3 Livestock fodder intake for meat (kq/dav) ' 2.710E+01 a 6.800E+01 3 --- 3 LFI5 19 3 Livestock fodder intake for milk (kaldav) 3 6.325E+01 3 5.500E+O '1 --- ' LFI6 19 3 Livestock water intake for meat (L/dav) ' 5.000E+01 35.000E+01 3 ' LWI5 19 3 Livestock water intake for milk (L/dav) ' 6.0005101 ' 1.600E+02 ' ' LWI6 19 3 Livestock soil intake (ka/day) ' 5.000E-01 5.OOOE-01 3 --- LSI 19 3 Mass loading for foliar deposition (C/m**3) ' 1.OOE-04 3 1.OOOE-04 3 --- ' MUFD 19 ' Depth of soil mixinfc laver (m '1.500E-01 ' 1.500E-01 ' --- 3 DM a 19 3 Depth of roots (m) 39.000E-01 3 9.000E-01 3 --- DROOT 19 3 Drinkinc water fraction from around water ' 1.OOOE+00 '2.OOOE+00 ' --- 3 FGWDW 19 3 Household water fraction from qround water ' not used 3 l..OOOE+00 ' --- 3 FGWHH 19 3 Livestock water fraction from cround water a 1.000E+00 ' 1.000E+00 ' --- ' FGWLW 19 3 Irrieation fraction from qround water 1.OOOE+00 ' 1.OOOE+00 ' --- ' FGWIR 3 3 3 9B 3 Wet weicht crop vield for Non-Leafv (ka/m**2) 3 7.000E-01 ' 7.000E-01 3 3 YV(1) 9B ' Wet weiqht crop vield for Leafv (kalm**21 A 1.500E+00 3 1.500E+00 ' -- 3 YV(2) 9B ' Wet weiaht crop yield for Fodder (ka/m**2) a l.lOOE+OO ' l.100E+00 ' --- s YVC3) 9B ' Growine Season for Non-Leafv (years) a 2,500E-0 3 1-700E-01 0 3 --- TEl) 9B 3 Growinq Season for Leafv (years) ' 1.230E-01l' .;2.500E-Ol ' --- 3 TEt2) 9B 3 Growing Season for Fodder (vears) ' 1.500E-01 ' 8.000E-02 ' - TE(3) 9B ' Translocation Factor for Non-Leafy ' l.OOOE-01 ' 1.OOOE-01 ' --- 3 TIV(1)

Appendix l Page! 18

SRAD, Version 6.0 To Limit = 0.5 vear 02/20/2001, 14:03 Paae 6 mmary : Braidwood with NRC Recomended Default Parameters File; BRAID1.RAD Site-Specific Parameter Summary (continued) 3 User ' Used bv RESRAD a Parameter nu 3 Parameter J Input . Default .2.(If-different-from-!user-input) 3 Name 9B 3 Translocation Factor for Leafv 3 9B ' Translocation Factor for Fodder TIV(2) £ 1.OOOE+0O1+00 ' - ' TIV(3) 9B ' Drv Foliar Interception Fraction for Non-Leafv ' E 2'500E-01 2.500E-O1 ' RDRY(1) 9B'3 Drv Foliar Interception Fraction for Leafy 32.500E- 2.l-2j500E-01 '

9B 3 Drv Foliar Interception Fraction for Fodder RDRY(2) 3 2'.'500E-01 .3 2.500E-01 ' - RDRY(3) 9B ' Wet Foliar Intercention Fraction for Non-Leafv: ' 2-500E-01 3'2.500E-01 a -

9B'.3 Wet Foliar Interception Fraction for+ Leafv RWETt1l 3 2.`500E-01 :f 2:.500E-01' --- 3 RWET(2) 9B ' Wet Foliar Interception Fraction 'for Fodder 3 2'500E-01 3 2.500E-01 ' --- 3 RWET(3) 9B 3 Weatherinq Removal Constant for Veqetation ' 2.000E+01 3 2.OOOE+01 3 3 3

--- ' WLAM 3 . 3 4 3 C-12 concentration in water (a/cm**3) ' not used ' 2.000E-05 I 4 3 C-12 concentration in contaminated soil (q/q) 3 C12WTR 3 not used ' 3.000E-02 3 --- 3 C12CZ 4 3 Fraction of veqetation carbon froni soil 3 not used '_2.000E-02 3 4 3 Fraction of veqetation.carbon from air --- 3 CSOIL 3 not used J 9.800E-01 33 4 3 C-14 evasion laver thickness in soil (m)

--- 3 CAIR

' not used 3.OODE-01 ' --- 3 DMC 4 3 C-14 evasion flux rate from soil (1/sec) 3 not used ' 7.000E-073 3 EVSN 4 3 C-12 evasion flux rate from soil (1/sec) 3 not used 3 1.OOOE-10 3 4 3 REVSN 3 Fraction of grain in beef cattle feed ' not used ' 8.000E-01 ' 2 AVFG4 4 3 Fraction of crain in milk cow 'feed ' not used 3 2.000E-O1 4 3 PCF correction factor for qaseous forms of C14 3 AVFG5 3

' not used ' 1.234E+02 ' --- ' C02F 3

OR ' Storaqe times of contaminated foodstuffs (davs): 3 OR 3 Fruits, non-leafv veqetables, and crain ' 1.400E+01 ' 1.400E+01 ' -

OR 3 Leafv veqetables STOR T(1) 3 l.'000E+00 2a.OOOE+OO ' 3 STOR T(2)

DR ' Milk 3 1.000E+00 ' 1.000E+00 '

DR 3 Meat and poultrv STOR T(3) 2.000OE+01 32.000E+01. 3 ' STOR T(4)

DR 3 Fish a 7.000E+00 ' 7.000+00 DR

--- STOR T(5)

' Crustacea and mollusks ' 7.000E+00 2 7.000E+00 ' STOR DR ' Well water T(6)

I 1.000E+00 3.1.'000E+00 '3 STOR T(7)

DR 3 Surface water 3 1.000E+00 1.000E+00 '

DR 3 Livestock fodder STOR T(8) 3 O.OOOE+00 4.500E+O1l ' --- STOR T(9) c 3 3 a 21 3 Thickness of buildine foundation (m) 3 not used ' 1.500E-01 -

21 3 Bulk densitv of buildinc foundation (q/cm**3) 3 FLOOR1 c 3 not used 3 2.400E+00 3 --- 3 DENSFL 21 ' Total porositv of the cover material ' not used ' 4.000E-01 3 ---

21 3 Total porosity of the buildinq foundation 3 TPCV

' not used 3 1.000E-01 . --- 3 TPFL 21 3 Volumetric water content of the cover material 3 not used 3 5.OOOE-02 3 21 3 Volumetric water content of the foundation

. I PH20CV q not used ' 3.000E-02 ' 3 PH20FL 21 3 Diffusion coefficient for radon qas (m/sec): a 21 3 in cover material 3 .

not used 3 2.000E-06 --- 3 DIFCV 21 3 in foundation material ' not.used 3 3.OOOE-07 3 21 3 3 DIFFL in contaminated zone soil . 3 not used 32.000E-06 ' 3 DIFCZ 21 3 Radon vertical dimension of mixina (m) 3 not used 3 2.OOOE+00 3 21 3 Averaqe buildinq air exchanqe rate (1/hr) 3 HMIX

' not used 3 5.000E-O1,' --- ' REXG 21 3 Heicht of the buildinq (room) m) not used: ' 2.500E+00 ' ---

21 3 Buildinq interior area factor 3 HRM

' not .used' - 'r.OOE+O' FAI .

21 ' Buildine deoth below around surface Cm) 3 not3used-l.000E+00 3 ---

21 3 Emanatinq power of Rn-222 aas 3 1 DMFL "

'not used ' 2.500E-01 I - EMANA (1) 21 3'Emanatinq power of Pn-220 uas 3 not used 3 1.500E-01 3 3 3 3 EMANA(2) 3 33 rL ' Number of araphical time points ' 32 ---

rL 3 Maximum number of Integration points for dose --- NPTS

' 17 3 LYM.AX Appendix 1 Page ( 18

SISRAD, Version 6.0 T( Limit - 0.5 Vear 02/20/2001 14;03 Pace 7 inmary Braidwood with'NRC Recomended Default Parameters File: BRAIDl.RAD Site-Specific Parameter Suzmnary (continued)

' User . Used'by RESRAD 3 Paxameter

!nlu Parameter ' Inpout 3 Default (if different from user input) 3. Iame

'Ift~~im~~ntwirf f ff+/-frffSiwfI~ff f f i~f tf if ff f f f f Tf f ff f ifffffif ff ffiff tf ff f f f ff f fiffIf tff tf f ifif ff f fff Summary of Pathway Selections Pathwav 3 User Selection 1 -- external qammna 'active 2 -- inhalation (w/o radon) active 3 -- plant ingestion active 4 -- meat incestion 3 active 5 -- milk ingestion active 6 -- aquatic foods ' active 7 -- drinkinq water ' active 8 -- soil ingestion active 9 -- radon .3 supressed Find peak nqthwqY,4 fffif3tffff oses c aive cfi feffft tftitt f~aifiaffwf fit Appedix Pag . 1 u

3SRAD, Version 6.0 Ta Limit = 0.5 Vear . 02/20/2001 14:03. Paqe 8 nmmary : Braidwood with NRC Recomended Default Parameters File: BRAID1.RAD Contaminated Zone Dimensions Initial-Soil Concentrations, vCi/ci Area: 3600.00 sQuare meters Co-60: . 1:800E-01 W

Thickness; 0.15 meters Mn-54 1.200E-01 aver Depth: 0.00 meters Total Dose TDOSEWt), mrem/vr Basic Radiation Dose Limit- 25:mrem/vr.

Total Mixture Sum M(tl Fraction of BasicflosehLimit Received at Time (t) t Cvears): 0.OOOE+00 1.OOOE+00 3.OOOE+00 1.OOOE+01- 3.000E+01 1.000E+02 3.OOOE+02 1.OOOE+03 TDOSEtt; 1.225E+00 9.951E-01 7.148E-01 2:686E-01 1.715E-02 8.695E-07 0.OOOE+00 0.000E+00 M(t): 4.901E-02 3.981E-02 2.859E-02 1.074E-02 6.860E-04 3.478E-OB 0.OOOE+00 0.OOOE+00

.ximum TDOSE(t): 1.225E+00 mrem/yr at t = O.OOOE+00 years N

N a

Appendx I Page 8 18

ESRAD, Version 6.0 T(< Limit = 0.5 vear 02/20/2001 14:03 :Paae 9 ummary : Braidwood with NRC Recomended'Default Parameters File: BRAID1.RAD Total Dose Contributions TDOSE(ipt) for Individual.Radionuclides (i).and Pathways (p)

As mrem/yr and Fractlon of Total-Dose At t = 0.OOOE+0O.years Water Independent Pathways (Inhalation excludes radon)

Ground Inhalation Radon Plant .Meat Milk Soil:

adio- IIKAXUDUU ;flAA AjA i _______ _ AAflkL

__ f uclide mrem/vr "fract. rnrem/vr fract. ' mrenm/vr -',,f ract.' e 'mrem/vr fr t fracm frac /Vr

=AAA kKAKAM AAh=~i~ AAIAAJUAA= A JA'4UYK=: h* h*KA Xjg- AAA= AAAJAAA o-60 1.040E+00 0.8492' 7.670E-08.0.0000 :0.OOOE+ OZOOaOOO 8.0358-03 0.0066'5.06E-.03-0.0041- 2828E-03.0.0023 .6.326E-05 6 48 0 5 0 00 0 rn-54 1 672E-01 0.1365 1,139EIQP 0.0000 Q.oooE+00. 0o0000 .50E-03 0 '0092 §ip- - 0 5 0-5 E 0

+/-+/-+/-t+/-f Hxifzttft f+/-+/-+/-t+/- fi+/-ftttII Ititfif fft~tffffft+/- tut  ; lIffti iti'titi +/-, I ',utu +/-ffiti+/-tititif +/-t+/-+/-+/-tit otal 1.208E+00 0.9857 7.784E-OB 0.0000 0.000+E00 0.0000 9.542E-03 0.0078 5.081E-03 0.0041 2.886E-03-0.0024 6.643E-05 Total Dose Contributions TDOSE(ip.t) for Individual Radionuclides (i),and Pathways (p)

As mrem/yr and Fraction of Total Dose At t= 0.000E+00.years Water Dependent'Pathways Water Fish Radon Plant Meat Milk All Path Idio- I M W iclide mrrem/vr fract. em/vr fract mrem/vr -fract.. mrem/vr' fract. mzmyr' fract. mrem/vr 'fract. rnrem/vr M.AAWUAAA A AU A XKAAAAh AkA. ARAXAXm AAA=,22:';UUUUUU.4 '2UU4AAt. A X

)-60 O.OOOE+00 0.0000 O.OOOE+00 0.0000 0.OOOE+.000.'0000 O.OOOE+00O.'0000 .O.OOOE+00:0.0000 0.000E+00, 0.000 1.0564+00 l-54 O.OOOE+OO OQ000 0 O.OOOE+OO.O.OOOO O OOE+02O.0000:.QO;OOOE+OO'Q.'OOOO' O.OOOE+OO ~O.OOOO0 :'OOOOE+OO:.O.OODO 1.6888-01

.fitf fffftftif fiffIf ffIf~ffff Iffif fdofifiIIt f fff ff hffIItfI!fififf i If fffft`-f ifft ;f fffff ItfffIt f fif .iff58-00

tal O.OOOE+00 0.0000 O.OOOE+00 0.0000 O.OOOE+00 0.0000 0.000E+00 0.0000 O.OOO+000.0000 0.00

.OOOE+0 0.0000 1.225E+OO lum of all water independent and dependent.pathways.

C C

4 Appendix I Page 9 ,18

ESRAD, Version 6.0 To Limit = 0.5 vear 02/20/2001 14:03 Paae 10 ummary Braidwood with NRC Recomended Default Pararmeters File: BRAIDl.RAD Total Dose Contributions TDOSEli,p,t) for Individual Radionuclides ti) and Pathways (p)

As mrem/yr-and 'Fraction of Total'Dose At t G lOOOE+00 . years Water Independent Pathways (Inhalation excludes' radon)

Ground Inhalation - Radon Plant  :-Meat '-Milk Soil adio- 1 AAAAAkM~

aclide mrem/vr *fra "trem/vr'-r'fract.

mrem/vr:; fract. mrem/vr fract. ' mrem/vr 4AA=UIS A ~ ~U~

AAA3=AUUAAAAA~,~hhh* ~ U~

v-60 9.074E-01 0.9118 6.660E-08'0.0000'- 0.OOOE+00,0.'0000.-6.978E-03,"0.0070 ';4.401E-03fv0"0044 2. 456-03C0.0025 5.'494E-05 l-54 7.316E-02 0.0735 4.960E-10 0.0000 O.OOOE+Oz 0 O0. 00 6.562E-04;0.^0007 5.943E-06 0.0000 2.517E-05 0.0000 1.37-7E-06 ffff f Iffiftt ffififf fffffiff +/-I;x+/- +/-f ft f+/-+/-+/-i i+/-+/- fiftI I+/-ff+/-+/-fiffif+/--ff+/-f+/-+/- 5f ff32-0i Dtal 9.806E-01 0.9854 6.710E-08 0.0000 O.OOOE+00'0.0000 7.634E-03'0.0077 '4.'407Z-03 0.0044 '2.481E-03-0.0025 5.'632E-05 Total Dose Contributions TDOSE(i.pt) for Individual Radionuclides UiY and-Pathways (p)

As mrem/yr'and Fraction of Total'Dose At t = 1.000E+00 years Water Dependent-Pathways Water Fish IRadon Plant Meat Milk All Path tdio-LClide mremfvr fract. mrem/vr fract. mrem/vr fract. mrem/vr';fract. *mrem/vr fract. *mrem/vr 'fract..

'AAAA AAA=.

0-60 0.OOOE+OO 0.0000 0.OOOE+00 O.0000 0.OOOE+000.0000 O.OOOE+OO;O.OOOO 0.0000E+Oi 0.0000 9.213E-01 L-54 0.000E+00 0.0000 0.OOOE+O00 0.0000 0.OOOE+OO'0000 O.OOOE+00 '0.0000 0 OOOE+00 0.0000 7.384E-02 0.OOOE+00 0.0000 t1tt tt+/-+/-+/-+/-t+/-f+/-t+/- +/-+/-+/-+/-+/-t+/-f+/- tit ut. ftfItfflf +/-+/-ttffU fft tufuit' 9.951E-01 ital 01000E+00 0.0000 O.OOOE+00 0.'0000 O.OOOE+00 0.0000 O.-OOOE+OO 0.'0000 0.OOOE+000,0000 um of all water independent and dependent pathways.

C LA C

4.

Appendix 1 Page 10 :8

XSRAD, Version 6,0 Tq Limit = 0.5 Vear 02/20/2001 14:03 Paae 11 C

• unmary : Braidwood with NRC Recomended Default Parameters File: BRAID1'.RAD z

Total Dose Contributions TDOSE(i,p,t) for Individual Radionuclides (i).and Pathways (p)

As mrem/yr and Fraction of Total Dose At t 3.000EI-00 years C Water Independent Pathways (Inhalation excludes radon)

Ground -Inhalation Radon ..Plant Meat 'Milk adio- Soil uclide Rm.rem/vr .fract. mrem/vr fract. racemyr rem/vctracvxnre/vr emart. ractm

/ f ier/yr o-60 6.902E-01 0.9656 5.022E-08 0.0000D -0OOOE+000.0000"'5.262E-03 0.00774' 3'.318E-030.'0046 1.852E-0300.0026 :4.43E-05 n-54 1.400E-02 0.0196 9.406E-11 0.00M0 0.'OOOE+00O.000Q 1.244E-04,0.0002 .1.127E-06 0.0000 4.773E-0G.0.0000 2.612E-07

+/-t+/-iffft+/- 0.9852 tfifft ffffIII0f tuu +/-+/-if+/-+/-iftt fii 5.031E-08 0ff.f0 tt ii flit titi ffffffO+I0'00f.000'.386E-03 +/-111 HIM 0.0075 +/-i+/-+/-+/-+/-i+/-30.00'iffift. t+/-i+/-+/-+/-Iiti +/-+/-lt otal 3.320ftE-0 tI56f fEf0.0fff 1fffffffI 7.042E-01 0.9852 5.031E-08 0'.0000 O.OOOE+OO rO'.OOOO 5'.386E-03 0.0075 3.320E 0.0046. 1:856E-03 ,0.0026 4-.169gE-05 Total Dose Contributions TDOSE(i. ,t) for Individual Radionuclides (i) and Pathways (p)

As mrem/yr' and Fraction of Total' Dose At t ='3.OOOE+00 years Water Dependent Pathways Water Fish Radon Plant Meat Milk All Path idio-aclide mrem/vr 'fract. mrem/vr fract. mrem/vr fract. mrem/vr fract. mrem/vr/:rfract.

AAAAKX 'mrem/vr :fract. mrem/vr I-so AKA=U: AAAAUA= AkUkAA JL =AA1AJUUAAM AAAAA1AA= , AUAAA~

)-60 0.OOOE+00 0.0000 O.OOOE+00 0.0000 0.OOOE+00 0.000 O.OOOE+00 0.0000 0.OOOE+00'O.0000 0.OOOE+00O.0000 7.007E-01 i-54 0.OOOE+OO 0.0000 0.OOOE+00 0.0000 0.00OE+00 0.0000' O.OOOE+00 0.0000 0.OOOE+00 0.0000 0.OOOE+00'0.0000

.fftff t+/-+/-ifttff ftftfi 7 13E-02 f f+/-ift f+/-f tfIftf 0.OiOOE+ ft00. 0000 it 0fffff 0 0f0000 0tfffO O - 0 00fff ital O.OOOE+00 0.0000 O.OOOE+00 0.0000 O.OOOE+00,0.0000 O.OOOE+OO 0.0000 O.OOOE+O O0.0000 O'.OOOE+OO O.OOOO 7.148E-01 lum of all water independent and dependent pathways.

C

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Appendix 1 Page 11 18

'SRAD, Version 6.0 TO< Limit = 0.5:Vear. 02/20/2001 14:03 Paae 12 umary Braidwood with NRC Recomended Default-Parameters. "File:..BRAID1.RMAD Total.,Dose Contributions ,TDOSE(iptv):for.,'Individu'alE.Radionuclides (i) and Pathways (p)

As nrem/yr :and Fraction of Total'"Dose Att = l.000OE+03 years Water Independent.'Pathways- (Inhalation -excludes'ra.don)

Ground-. Inhalation Radon Plant Meat -Milk Soil

.dio-' AAAIXQ4AAAAAAkA. AU=SSl~l .KAKAI

.clide mrem/vr fract. mrem/vr 'fract. mrem/vr fract' mremr/vrfract. mrem/vr;fract. nmrem/vr "fract.. : mrem/vr A A2 . AKA=2- }AAU AUAA=: )' - AA .k-A AAiUM

-60 2.646E-01 0.9853 1.867E-08.0.0000 O. OOE+00;'0.00000 '1'956E--030;.0073. 1.233E-03 0.0046 .882E-04 0.0026 1.540E-05

.-54 4.293E-05 0.0002. 2.789E-13 0.0000 0.O00O+00 .000003.690Q-07'0.'0000 3.342E-09:0.0000 -1.155-08 0.0099 7.744E-10 fIft hfiff itt fffitf i+/-ftfftiIt~ tfIfi ttttitt' :iitif t Ii t tI:iftff ufi tt tif+/-fI'Iif+/-f fff ftftttt'ffiiIi itiifffift

'tal 2.647E-01 0.9655 1.867E-08 0.0000 O.OOOE+00'0.0000 1'.956E-03 0.0073 1.233E-03 0.0046 6.882E-04 0.0026 1.540E-05 Total Dose Contributions TDOSE(i,p,t) for Individual Radionuclides li) and Pathways (p)

As mrem/yr and Fraction of Total Dose At t 1.OOOE+0l years Water Dependent Pathways Water Fish Radon Plant Meat Milk All Path dio- ~AAAA2AAAJA clide mrem/vr fract. mrem/vr fract. mrem/vr fract. mrem/vr fract. mrem/vr "fract. :mrem/vr. fract. mrem/vr AAAAA=UUIA&

-60 O.OOOE+00 0.0000 O.OOOE+00:0.0000 0.OOOE+00 0.0000 0;OOE+00.0O.0000, 0O.OOOE+00..0.0000 0.OOOE+00O.OOOO 2. 685E-01

-54 0.OOOE+00 0.0000 :900OE+00 0.0000 0.OOOE-400 0.0000..OOOE+0070.0000 O.OOOE+00.0.0000 4.332E-05 0ffOffOffO 0tf0f 0fOffOffOfff

+ff ftOi-0ffff 0ff.00 'f0fiff.fOf0f0f00 ffffffIff ffffff, ifffItfItI tal O.OOOE+OO 0.0000 I0.000E+00 0.0000 O.OOOE+OO'O.DOOO+ O.OOOE+OO 0.0000 O0.OOOE+OO 0.0000 O.OOoE+0 .0.0000 2.686E-01 am of all water independent and dependent pathways.

C u2 C.

N a1 I

Appendix 1 Pageo 1 18

ESRAD, Version 6.0 Tc Limit 0.5 vear 02/20/2001 14:03 Paae 13 ummary : Braidwood with NRC Recomended Default Parameters File: BRAID1.RAD Total Dose Contributions TDOSE(i,p,t) for Individual Radionuclides (il.and Pathways (p)

As mrem/yr'and Fraction. of Total"Dose.Att'= 3.000E+0l years Water Independent Pathways (Inhalation excludes radon),

Ground. Inhalation' Ra don Plant! meat, Milk :Soil adio- RISJO btiliAn OI]g s~LAk- 22UJUUt'tU@U2,2A iclide mrem/vr fract. ;rem/vr fract. m/vr.: fract. -=;mrem/vr fract. mrem/yr Sfract. vr-fra AAUA= AAM A4X UU.AWW2A 'AAAA AAAAAA=2U .AUU u" 'AAAJUU=A2, AAA MU=JUiAi

-60 1.692E-02"0.9868 1.08SE-09 0.'0000- 0.OOOE+00, 0.0000 l.140E-04 0'.-0066. 7.189E--05 0'.0042 4'.'011E-05'0.0023 8.973E-07 1-54 2.799E-12 0.0000 1.64$E-20 0.0000 0.OOOEs+00Q.0000 -2.177E-14;0.0000' 1'.972E-1`0.`0000 8.349E-16-'0.000 4 .56BE-17 t+/-f+/-f+/- tf+/-Itff tIf if ff ffIlfff t itI'ff fftfl iffffi fI+/-ff1+/-1 ::+/-+/-+/- tif1ff1+/- tIitfiff +/-11+/-f I' +/-+/-+/- tiiifffI iIttifitt

)tal 1.692E-02 0.9868' 1.0BE-09 0.0000 O.OOOE+00 0.0000 1.1405-04 ;0.0066 7.189E-05 0.0042 4.011E-05 0.0023 8.973E-07 Total Dose Contributions TDOSE(i,oft) for Individual. Radionuclides (il and Pathways (p)

As mrem/yr and Fraction of Total Dose At t = 3.000E+01 years Water Dependent Pathways water Fish Radon Plant Meat Milk All Path Ldio- AKAAKKXAAAAAA =

iclide mrem/Yr' fract. mrem/vr fract. mrem/vr fract. mrem/vr fract. , mrem/vr fract.. mrem/vr fract. mrem/vr J~AA= AAAA AA.MAA ww .AA. *.A.AA b- ~.-Ajkk AAAAAAAM

'-60 O.OOOE+00 0.0000 0.0005+00 0.0000 0.OOO+OO 'O.0000 .0.000E+00 O.0000 0.000E+000'.O000 .O.OOOE+OO 0.0000 1.715E-02

-54 0.QOOE+O0 0.0000 O.OOOE+OO00.0Q000 0.OOOE+000.0000 O.'OOOE+00 0.0000 0.000E+000O. 0000- -. OOOE+00-.0.0000 2.B822E-12 fftf+/- fIlttff ff1ft1 tiff lff IttItII tiff fiff fI fft ffiuffi+/- I tIttI uuftuitui fitftff :itffi tt f tt1ttf tfftfff tt tal O.OOOE+00 0.0000 0.0005+00 0.0000 O.OOOE+00 0.0000 O.OOOE+00 O.0000 0.OOOE+OO 0.0000 0.OOOE+00 0.0000 1.715E-02 um of all water indeperident and dependent pathways, C

CA 2

a m

E!

t:

Appendix Page 13 18

ESRAD, Version 6.0 T<< Limit = 0.5 Vear 02/20/2001 14:03 Pane 14 ummary : Braidwood with NRC Recomended Default Parameters File: BRAID1.RAD Total Dose Contributions<TDOSE(i,p,t) for Individual Radionuclides. li) and Pathways (p)

As mrem/yr and Fraction of:'Total-Dose At t = 1.OOOE+02 years Water Independent:Pathways (Inhalation excludes-radon)

Ground Inhalation Radon Plant Meat Milk .Soil adio- I iclide mrem/vr: fract. mrem/vr' :fract.  :.mrem/vr;.'fract. .-rrem/vr frfract mr'fr act. mrem/vr:fract. mrem/vr AAWU= A. X:APh. S6 ' W'AAAAU2.

AA i-60 8.618E-07'0.9911. 3.70J.-14 0.0000 .:0.000E+00f~0.0000- 3.879E-09'.0.0045 *2..447E-09 0.0028:1.365E-09;.0.0016 :3.-053E-11 l-54 0.OOOE+00:0.-0000. O.OOOE+OO-O.OOOO  :.O:OOOE+OO.O.'OOOO.mO.OOOE+O 0.00 O.OOE+O 0.-0000 O.OOOE+00 0.0000 0.OOOEt00 tIffi+/- t+/-tiff+/-fi fiffff ffffff fffIfI MII f Iff-ff Ifftii f f tft! iifffftif ff ff it: f+/-Iif+/-+/-+/- ftfiiI +/-+/-fiff+/-i+/- t+/-fffft I~t+/-ff+/-tt 3tal 6.61SE-07 0.9911 3.701E-14 0.0000 o.o0oE+00 0.0000 3.879E-09 0.0045 2.447E-09 0.0028 1.365E-09 0.0016 3.053E-11 Total Dose Contributions TDOSE(i.D,t) for Individual Radionuclides .(i) .and Pathways (p)

As mrem/yr and Fraction of Total Dose At t = l.OOOE+02 years Water Dependent Pathways Water Fish Radon Plant Meat Milk All Path Idio-iclide mrem/vr fract. mrem/vr fract. mrem/vr fract. mrem/vr fract. mrem/vr fract. mrem/vr fract. mrem/vr AAAAIAM A~AAA AAAA MU= A2U, AJA2".AAAAAW,'AM= AA3UAM  : AAAi :

,-60 O.OOOE+00 0.0000 0.OOOE+00O0.0000 0.000E+00 .00000 _0.000E+00 .00000,O.OOOE+00:0.0000 O.OOOE+00-O.0000. 8.695E-07

-54 0.OOOE+0 0.0000 0.OOOE+0O 0.0000 O.OOORE+00 O.0000 0000E0000000 0.OOOE+ . .000OE+0 0.0000 0.OOOE+00 f f0if-f0f I0fttf fi+/-+/-+/-f+/-f t+/-fffff fitftfffI ftfif .+/-I+/-fff+/-f+/-+/-f fiffft If+/-I+/-ffft fiftit itIt fitfiififI fitf ttft f tal O.-OOOE+00 0.0000 O.OOOE+00 0.0000 0.OOOE+00 0.0000. O.OOOE+00 0.0000 O.OOOE+00 0.0000 0.000E+-0O0.0000 8.695E-07 um of all water independent and dependent pathways.

C

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Appendix I Page 14 18

!SRAD. Version 6.0 TV Limit = 0.5 vear 02/20/2001 14:03 Paae 15 anzary Braidwood with NRC Recomended Default Paxameters File: BRAIDl.RAD Total Dose Contributions TDOSE(ipt) for Individual Radionuclides (i) and Pathways (p)

As mrem/yr and Fraction of Total Dose At t 3.O0OE+02 years Water Independent Pathways (Inhalation excludes radon)

Ground Inhalation Radon Plant Meat' Milk Soil Ldio-iclide mrem/vr fract. fract.

mrem/vr mrem/vr fract. e r r mrem/vr fract. rnrem/vr fract. mrem/vr

'g

)- 60 O0.000,-i-o0 .0000 O.OOOE+OO0.000 DO O DOOA Ot:O O.OOE+OO4.00 00+0.0.OOO ~;~O O.OOOE+OO..O.:OOOO :QOOOE+00'-O'~.OOOO0 O.OOOE+OO

&-54 0.OOOE+O' O.0000 0.O0OE-I00 0.0000 0.00 00,.0000 0o.OOOE+00 0.00 oo .f000E+00.`0.0000 .0.000E+00 0.000 0.0.OOE+OO 21tiII Ital O.OOOE+00 0.0000 O.OOOE+OO 0.0000 0.OOOE+00:0.0000 0.000E+00 O.0000 O.OOOE+00 0.0000 O.'OOOE+00 0.0000 O.OOOE+00 Total Dose Contributions TDOSE(ip.t) for Individual Radionuclides i).and Pathways (p)

As mremlyr and'Fraction of Total Dose At t = 3.OOOE+02 years Water Dependent Pathways Water Pish Radon Plant Meat Milk All Path

-dio-clide mrem/vr fract. mrem/vr fract. mrem/vr fract. mrem/rr fract mrem/vr-fract. mrem/vr fract. mremn/vr AAA AAAA2U=, ~AAA1U A=AA~ k~ AAXAM:AK= AAA

'A AXA 2.A1 XUJLAAAA=

-60 O.OOOE+00 0.0000 0.OOOE+OO 0.0000 O.'OOOE+000.0000 O.OOOE+000.0000 O.OOOE+00: O.'0000 O.0O0E+OOO.0000 0. 000E+00

-54 0.OOOE+0 0.0000 O.OOOE+00 0.0000 0.OOOE+00 0.0000,'0.O00E4-00 0.'0000 0.000E+0-0 00000" O.00E+O00 0.00w 0.OOOE+00 tf ftf hitifftif HIM~ +/-Ififf fifitIf If+/-f4f+/-f+/-:+/-+/- fffl +/-t+/-+/-+/-

ffft+/-f fff if f ff 0t I fff itif ti 1f+/-+/- t1ff 1t+/-ftV 0fff.fff tal O.OOOE+OO 0.0000 O.OOOE+00 0.0000 0.000E+00 0.0000 O.000E+00 0,0000 O.OOOE+O0 O.0000 0.00O0E+00 0.0000 O.OOOE+OO um of all water independent and dependent pathways.

C C

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Appendix 1 Page 1' 18

ESRAD, Version 6.0 T(( Limit -0.5 Vear :02/20/2001 14:03 Pace 16 ummary Braidwood with NRC Recomended.Default Parameters File: BRAID1.RAD Total Dose Contributions.TDOSZUi,P,t) for Individual-Radionuclides (i) and.Pathways (p)

Asnmrem/yr andFraction of Total ose.-At t= 1.OOOE+03 years Water Independent Pathways (Inhalation excludes radon)

Ground Inhalation Radon Plant Meat Milk Soil adio- flflA2iAAfA hiA 2KA .

uclide mrem/vr fract. mrem/vr fract. mrem/vr fract. mrem/vr fractt, mrem/vr fract.. mrem/vr fract. mrem/yr AAAUUKh AA k .AAAAAA=!. AAA= AA2 A22A 1UA 2 AA A~AAAAAM D-60 O.OOOE+00 0.0000 .0.ooo+0. 0.000 0.o O.OOE+00.O..0000: .OOOE+OOO.00 O.OOOE+OO O.0OO O.000E+00. 0.0000 0.'

.000+00

.i-54 0.0005+00 0.0000 OOOOE40 0.00 0 O .0005+00 0.0000 0.OOOE+0010.0000 0;OOOE+009.O00q0 , 0 .OOOE+00-.O0000 0 .'000E+00 if ittitif iit+/-fi I+/-+/- fff+/-+/-tti it t +/- +/-1fftttfitf fffi ttzff i 1t+/- tf 'tittit +/-t1f+/-+/-+/-+/-t+/- :ift:r+/- +/- fi +/-+/-iff+/-.41 1t fIttIttff

.tal O.OOOE+00 0.0000 O.OOOE+00 0.0000 O.OOOE+00 O.OOOO O.OOOE+00O.0000 O.OOOE+tOO0.0000 0.000E+00.O000 0.000E+00 Total Dose Contributions TDOSE(iaD,t) for Individual Radionuclides (i) and Pathways (p)

As mrem/yr and Fraction of Total Pose At t 1.000E+03 years Water Dependent Pathways Water Fish Radon Plant Meat , ' Milk All Path idio- AKAAKA2Ai 2Alki AAK A AA. AA iclide mrem/vr fract. mrem/vr fract. mrem/vr fract. mrem/vr 0 fract. mrem/vr fract. -mrem/vr^: fract. mren/vr 1AX2AA Au ;a A kAA. xuA2uuuU MAAAJA: A.A'A. 'X'..

-60 O.OOOE+00 0.0000 0:.0O0E+00 0.0000 O.DOOE+O 0..O.0000. 0.000E+00:0.0000f 'O.OOOE+00. 0.0000 ;OOOOE+00 O.O000,. 0.000E+00 1-54 O.OOOE+OO .000O .0.000E+00 0.0000 0.000Z+00OO.000 0.oOOOE+D00i.0000; 0DoooE+00oo.00OO..00OE+0O0.1e0000 0.00Q+.00

.:00 I~~ t~ifIIff+00fff ->lf-tl-:IfiM iffIf ififf-+t t.fffffftff.f tf f f .:. 'ftf.f ff 'ttilffi ff ff:-

5tal 0.000E+00 0.0000 0.0005+0 00000 0.0OOD+OO 0'0000 0. OE 0 a0'-.OOO0 IOOO

. +w000.00OO000 0 0.0005+00  :

^.1umlof all water independent and dependent, patways.;; .  ? :er.; ,<. . ......

lf +/-fIff U

0 Appendix IPage If 18

SRAD, Version 6.0 T< Limit = 0.5 vear 02/20/2001 14:03 Paae 17 mmary : Braidwood with NRC Recomended Default Parameters File: BRAID1.RAD Dose/Source Ratios Suzmed Over All Pathways Parent and Progeny Principal. Radionuclide'Contributions Indicated

.rent Product Branch DSR(1,t) (mrem/vr)/(PC i/a) i) (1) Fraction* t= O.OOOE+00 l.OOOE+00 3.OOOE+00 1.OOOE+01t'.3.000E+01 1.0O0E+02-.3.000E+02 .1.000E+03 A' ;; ': A2'A2UzAAAA M.

• -60 Co-60 l.OOOE+00 5.869E+00 5.-118E+00 3.893E+00 1.4921+-00 9.528E-02 4'.831E-06 o0.'OOOz+aO O.OOOE+00

-54 Mn-54 1.OOOE+00 1.406E+00 6.154E-Ol 1.178E-01 3.610E-04 2.352E-11 1.289E-36 0.000E+O0 0.OOOE+00 ffff I iffffff tII+/-ffff tIffffifI ftffifffif f.+/-ttti+/-tf'fffiffII+/-f fIItfIft fffIfffi ff+/-iftfI+/- +/-ffffIIIII ranch Fraction is the cumulative factor for the iIt Drincipal:radionuclide dauqhter: ::CUMBRFMi). BRF(1)*BRF(2)* ... BRF(1).

e DSR includes contributions from.associated'(half-life 6 0.5.yr)daughters.

Sinale Radionuclide Soil Guidelines G(it) in pCi/q Basic Radiation Dose Limit - 25 mrem/yr clide (i}) t= O OOOE+OO 1.OOOE+00 3.OOOE+OO 1.OOOE+01 3.000E+o0 1.OOOE+02 3. 000E+02 1. OOOE+03

-60 4.260E+00 4 .884E+00 6.422E+00 1.'676E+01 2. 624E+02 5.175E+06 *1.131E+15 *1'.131E+15

-54 1,777E+01 44.063+01 2.123E+02 6.'926E+04 1.063E+12 *7* 7'44E+15 *7.744E+15' *7 744E+15 itjij ifitiifff ifIfftffI IfiIfftt tf+/-f~ifii fiftfift: f tfIIIIII t specific activity iim +/-iit Summed Dose/Source Ratios DSR(it) in C(mrem/vr)/'pCi/cfl and Sinnle Radionuclide Soil' Guidelines G(it) in.pCi/a at tmin = time' of minimum sinale radionuclide soil Quideline and at tmax = time of maximum total dose -'O.OOOE+00 years

lide Initial tmin DSRfi tmin) GHi.tmin) DSR(itmax) G(i~tmax)

Ci) p3Ci/QT (vears) . (pCi/q) ilA # AAAAJ

• 60 1.800E-01 0.OOOE+00 5.S69E+00 4.260E+00 5. 869E+00 4.260EOO

.54 1.200E-01 0. 00024-00 O.OfDtf ffOOfff }. 406E+00 1 777E201 1.406E+00 1.777E+01 c iftff fitiffffI iffft f itit Ht ffit Ififtifit tiffitiff z 0

0.

Appendix I Page 17 18

ESRAD, Version 6.0 T(< Limit = 0.5 vear 02/20/2001 14:03 Paae 18 ummary Braidwood with NRC Recomended Default Parameters File: BRAID1.RAD Individual Nuclide Dose Summed Over A11.Pathways Parent Nuclide and Branch Fraction. ndicated iclide Parent BRF(i) ,WOSE(it). mrem/vr (i) h t= 0.OOOE+00 l.000E+00 13.00E+0Ol,.000E+Dli;3.OOE+01: 1.OOOE+02 3.OOOE+02 1.OOOE+03 z-60 Co-60 1,000E+00 1.056E+00 9.213E-01: 7.007E-01. 2.685E-01 1,715E-02 8.695E 0.OOOE+00 0..000E+00

'i-54 Mn-54 1.00oE+00 1.68BE-01 7.384E-02 l.413E-02 4.332E,-Q5 2.B22E-12 0.00OE+00o 0.o000+00.0.000+ot 00 IF(i) is the branch fraction of the parent nuclide.

Individual Nuclide Soil Concentration Parent Nuclide and Branch Fraction Indicated iclide Parent BRF(i) S(iut), ICi/a j)) t= O.OOOE+00 1.O000E+00 3.000E+00 1. 000E+01 3.OOOE+01 1.OOOE+02 3.000E+02 1.000E+03

':A A A LAAAA.A A A .4 .A.2iUA=A 3-60 Co-60 1.000E+00 1.800E-01 1.574E-01 1.203E-01 4.695E-02 3.194E-03 2.623E-07 5.569E-19`0.OOOE+00 i-54 Mn-54 1.OOOE+0o 1.200E-01 5.261E-02. .0UE-092 3.149-0 2.26BE-12 1.8584-37 O.oo E+oo0 0o.oooE+o tF(i) is the branch fraction of the parent nuclide.

'SMAIN5.EXE execution time = 21,86 seconds Appendix I Page IF .18

Based upon the evaluation from Millennium Services, Inc. which provided a recommendation to post the area of soil contamination (due to Co-60) resulting from the vacuum breaker leak, those sample grids containing Co-60 were subdivided and sample.

The attached drawing indicates those grids sampled (Grids 3, 6, 13, and 35). Isotopic analyses of the samples are attached.

Grid 3 indicated one quadrant with Co-60 at a concentration of 2.93 IE-7 uCilg. Based on a quadrant size of 5m x 5m (depth of 6 inches) and a soil density of 1.6 gc'c, the total uCi of Co-60 is determined to be 1.78E+0 uCi, which is well below the limit of 10 uCi which would require radiological posting.

Additionally, in one quadrant of Grid 3, isotopic results indicated the presence of Mn-54 at a concentration of 1.838E-7 uCi/g and in one quadrant of Grid 13 isotopic results indicated the presence of Te-123m at a concentration of 5.818E-8 uCi/g. Neither of these isotopes was present in the initial grid samples. The activity of these quadrants, when added to the original totals, was evaluated against the posting criteria of IOCFR20.1902(e) and determined that posting of the spill area is not required.

C C/CE LOICGtCLRI pOompIlim j~Um nuLIcn

Sampling Grid N

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QUESTIONS FROM SENATOR DURBIN'S STAFF REGARDING THE DRESDEN STATION TRITIUM LEAK

Background

The licensee identified on August 30, 2004, through sampling of shallow wells on site that there were elevated levels of tritium in some locations. Onsite deep well samples were not at elevated levels nor were elevated levels of tritium identified offsite in either surface or well water samples. The source of the leakage was identified as a buried common suction line from the condensate storage tanks to the Unit 2 and 3 High Pressure Coolant Injection (HPCI) systems.

The leakage was largely confined to a small shallow area outside an adjoining plant building and has posed no environmental hazard. Tritium contaminated water was identified in the storm drain system but has not traveled offsite . Measurements at the storm drain outflow in onsite locations showed concentrations that were less than half of the EPA drinking water limit.

Although the condensate storage tank is the normal source of water to the HPCI systems, the torus is the safety related source of water and the HPCI systems for both units are currently aligned to the torus. Therefore, HPCI system function is not impaired by the line leak from the condensate storage tanks. No EPA tritium drinking water or NRC effluent release limits have been exceeded as a result of this leak.

The licensee started excavating on September 3, 2004, to attempt to identify the exact location of the leak. Difficulties with contractor expertise resulted in the inability to identify the exact source of the leak and personnel safety issues within the excavation. The licensee stopped work due to the second issue and hired a different excavation contractor. Due to difficulties with excavation near existing equipment and the inability to identify the exact location of the leak the licensee has decided to reroute a 75-foot section of the piping and abandon portions of the underground piping in place. The work is ongoing.

Question # 1:

What requirements exist for informing the public about leaks like this ?

Response

Title 10 of the Code of Federal Regulations (CFR) Part 20, "Standards for Protection Against Radiation," provide many of the reporting and notification requirements for radiological issues.

These requirements are contained in Subpart M, "Reports," which provides the reports of most radiological issues that NRC licensees are required to make to the NRC. 10 CFR Part 50 in 50.72, "Immediate Notification Requirements for Operating Nuclear Power Plants" and 50.73, "Licensee Event Reporting System," provide emergency notification requirements and those for reporting events that relate primarily to reactor operating conditions.

While the regulations in 10 CFR Part 20 include NRC notification requirements in case of releases of radioactive material above prescribed limits and for radiation doses to the public in excess of specified limits, the tritium leakage that recently occurred at the Dresden Station is not reportable to the NRC because none of the reporting thresholds were reached.

The licensee is required by their operating license to implement a program for radioactive effluent controls and for monitoring the potential impact of radioactive effluents on the

environment through a radiological environmental monitoring program (REMP). The REMP requires sampling of various environmental pathways including waterborne pathways at required intervals, which are to be analyzed for the presence of specified radiological constituents. Reporting levels for radioactivity concentrations in environmental samples are specified in the REMP and include reporting levels for tritium in water. Should the "reporting levels" specified in the REMP be exceeded, the licensee would be required to prepare and submit a report to the NRC that identifies the problem and defines its corrective actions. The problem would also be required to be reported to the NRC in the license's Annual Radiological Environmental Operating Report. The reporting level for tritium required by the REMP was not approached for this Dresden leak.

There are no requirements for licensees to directly inform the public of leaks or to inform the public of other radiological issues that may not otherwise be reportable to the NRC under 10 CFR Part 20 or Part 50. However, should licensees make required reports to the NRC, such reports are made available to the public (absent safeguards information) on the NRC's external web site.

Question # 2:

What NRC requirements are there for licensee's to fix, monitor and contain a leak (like this)?

Response

Should leaks occur, a licensee would be required by 10 CFR Part 20 (20.1501) to evaluate (i.e.,

monitor) the extent of the leak so as to assess its potential radiological hazard and assess its radiological impact. Following that evaluation, the licensee would be required to correct the problem to ensure the leak would not result in effluent releases or radiation dose to members of the public in excess of regulatory limits.

Question # 3:

What are the statistics on how often such tritium leaks have occurred at Dresden?

Response

The NRC does not maintain statistics or records of leaks that occurred at Dresden that are below the reportability criteria . However, if a leakage problem or other radiological issue was reportable under 10 CFR Part 20 or Part 50, the licensee's required report would be maintained along with the report of any NRC inspection that evaluated the issue.

The licensee has conducted a REMP since the early 1970s. Through this program, radiological

_impacts to workers, the public, and the environment are monitored, documented and compared to the applicable standards. As part of the NRC's inspection program, we routinely evaluate the adequacy of the licensee's REMP and ensure that the appropriate environmental pathways are sampled and analyzed as required. These inspections are currently performed on a biannual basis and have not identified significant problems with either the development or the implementation of the Dresden Station REMP.

The potential environmental impact of radiological releases from the plant are required to be summarized in the Annual Radiological Environmental Operating Report and the Annual

Radioactive Effluent Release Report. These reports are also reviewed as part of the NRC's routine inspection program. These reports have shown no discernable radiological impact on the environment from Dresden Station operations. These reports are also available for public review on the NRC's external web site.

Question # 4:

Will remediation of the dirt where the tritium leaked be necessary?

Response

No immediate remediation of the soil is required unless the licensee plans to relocate any of the soil that was excavated to repair the line leak to another location at their facility or dispose of the soil as radioactive waste. In that instance, NRC approval would be required under the disposal provisions of 10 CFR 20.2002. Should the licensee return the excavated soil back into the excavated area or otherwise have not disturbed the soil, then the licensee would be required to maintain a record of any contamination in and around its facility for future decommissioning purposes pursuant to 10 CFR 50.75(g). Dresden plans to opt for the 50.75(g) methodology.