ML20140D667

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Submits Response to Violations Noted in Insp Rept 50-155/97-02.Corrective Actions:No Worker or Loading Dock Area Contaminations & Radiation Protection Surveyed Remaining Lead Sheets
ML20140D667
Person / Time
Site: Big Rock Point File:Consumers Energy icon.png
Issue date: 06/05/1997
From: Addy R
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
50-155-97-02, 50-155-97-2, NUDOCS 9706110037
Download: ML20140D667 (12)
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A CMS Energy Company Big Rock Potnt Nxtear Plant Kennen P. Pneuws 69 Plant GeneralManager l June 5. 197 l Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 i DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT - REPLY TO A NOTICE OF l VIOLATION - NRC INSPECTION REPORT 97002.

During a routine NRC inspection conducted from January 18, 1997, through March 12, 1997, four violations of NRC requirements were identified and forwarded by letter dated April 23, 1997.

The first violation involved a loss of control of radioactive material outside l the radiologically controlled area. The second violation resulted from not performing inspections on two electrical penetrations as required by Technical Specifications. The third and fourth violations document the failure to perform a written safety evaluation for a potential unreviewed safety question (USQ) involving containment post accident temperatures and the failure to reflect the results of an analysis of containment temperatures in an update to the FHSR.

Consumers Energy Company agrees with the first and second violations as stated. With respect to the third and fourth violations, additional consideration is requested. Supplemental information that has been retrieved since the time frame of the inspection has been provided. A 15 day extension was granted by the Region III Branch Chief in order for the Big Rock Point staff to assimilate and docket the information.

Pursuant to the direction provided in the report, find attached a Reply to the Notice of Violation. The proposed corrective actions are intended to address the concerns identified by the violation, and to prevent recurrence of the violations.

N Robert J Addy Plant Manager

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CC: Administrator s,qegion III, USNRC l NRCRe{ip(pFIndpector-BigRockPoint ATTACHMENT 9706110037 970605 lljl ll{lll,0llllllllll*llll a c

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ATTACHNENT CONSUNERS ENERGY CONPANY BIG ROCK POINT PLANT DOCKET 50-155 REPLY TO A NOTICE OF VIOLATION INSPECTION REPORT 97002 Submitted June 5, 1997 1

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4 REPLY TO A NOTICE OF VIOLATION - NRC INSPECTION REPORT 97002 1

, NOTICE OF VIOLATION

Consumers Energy Company Docket No. 50-155
Big Rock Point Nuclear Plant License No. DPR-6 8

During an NRC inspection conducted from January 18, 1997, through March 12, 1997, four examples of violations of NRC requirements were identified. In l accordance with the " General Statement of Policy and Procedure for NRC 1

Enforcement Actions," NUREG-1600, the violations are Ifsted below:

1. Technical Specification 6.11 requires that procedures for radiation

! protection shall be prepared consistent with the requirements of 10 CFR i

Part 20, and shall be approved, maintained and adhered to for all operations involving personnel radiation exposura.

Administrative Procedure 5.11, " Radioactive Material Control," Revision 13, Step 5.4.1.1(e) requires that storage areas have positive access control (locks, fences, stationed personnel or barricades) maintained by the Chemistry and Health Physics Department.

l Contrary to the above:

l l Between January 27-28, 1997, the licensee identified the presence of

stored contaminated lead outside the radiologically controlled area, an 1 area that did not have positive access
ontrols.

l This is a Severity Level IV violation (Supplement IV).

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2. Technical Specification 3.7(c) requires that "Each reactor shutdown for l refueling, but'in no case at intervals greater than two years, all electrical and accessible piping penetration nipple welds be visually examined for evidence of corrosion, cracking or deterioration."

Contrary to the above, on January 15, 1997, the licensee discovered that two containment electrical penetrations (H-115A and H-1158), previously considered inaccessible, were in fact available for inspection but had not been inspected since installation in March 1985.

This is a Severity Level IV violation (Supplement IV).

3. 10 CFR 50.59 requires, in part, that changes in the facility, as described in SAR (FHSR at Big Rock Point), shall not be made without prior NRC approval unless it is determined that the change does not represent an unreviewed safety question (USQ), and a written safety evaluation documents the bases for this determination.

Contrary to the above, on February 7,1985, the licensee failed to perform a written safety evaluation to document the basis for determining that a USQ did not exist when it was determined that containment post accident temperatures would exceed those identified in the FHSR.

This is a Severity Level IV violation (Supplement I).

4. 10 CFR 50.49 requires, in part, that the licensee establish an equipment qualification program which must include and be based on the time-dependent temperature at the location of the equipment following the most severe design basis accident for which this equipment is required to remain functional.

j REPLY Y0 A NOTICE OF VIOLATION - NRC INSPECTION REPORT 97002 2 10 CFR 50.71(e) requires, in part, that FSAR (FHSR at Big Rock Point)

. ' updates shall contain all changes necessary to raflect analyses prepared pursuant to NRC requirements.

Contrary to the above, on December 22, 1989 the Iicensee submitted a FHSR update which failed to reflect the results of analysis on time-dependent temperatures inside containment following design basis accidents, which had not been prepared pursuant to 10 CFR 50.49. l l This is a Severity Level IV violation (Supplement I). l l

l Consumers Energy Company's response is provided below.

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! REPLY TO A NOTICE OF VIOLATION - NRC INSPECTION REPORT 97002 3 Violation 97002-01 Technical Specification 6.11 requires that procedures for radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20, and shall be approved, maintained and adhered to for all operations involving personnel radiatton exposure.

Administrative Procedure 5.11, " Radioactive Material Control," Revision 13, Step 5.4.1.1(e) requires that storage areas have positive access control (locks, fences, stationed personnel or barricades) maintained by the Chemistry and Health Physics Department.

Contrary to the above:

Between January 27-28, 1997, the licensee identified the presence of stored contaminated lead outside the radiologically controlled area, an area that did not have positive access controls.

Consumers Energy Company agrees with the violation as stated.

I. Reason for the violation On January 24, 1997, lead sheets were transferred from the radioactive waste storage facility to the machine shop loading dock. The lead being transferred consisted of newly purchased lead stacked on a pallet of previously used lead.

The lead sheets were going to be used for fabricating shielding for radwaste filters.

Prior to releasing the lead sheets for transfer, the Radiation Protection (RP) supervisor was advised by RP personnel that the lead had been radiologically surveyed. No radiation / contamination had been discovered. In addition, the RP supervisor performed a masslin smear of accessible surfaces (top and sides of the stack). The masslin survey confirmed that those accessible surfaces were free of loose contamination. The RP supervisor released the lead sheets for the filter work.

The RP supervisor expected the lead sheets would be moved from the loading dock directly into the Radiologically Controlled Area (RCA). He called maintenance to followup, and was told that the lead sheets had been brought in off the loading dock. However, only the lead sheets needed had been brought into the RCA. The pallet with the remaining lead sheets remained on the loading dock.

On January 28, 1997, a repair worker began re-arranging the lead sheets one-by-one to create a path so that a compressed gas bottle could be moved across i the loading dock. He discovered a radioactive materials sticker on one of the lead sheets, stopped what he was doing, and notified Radiation Protection personnel.

! The root cause of this event was determined to be a failure to perform an adequate radiological survey of material being released from the Radwaste i facility.

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II. The corrective steos that have been taken and the results achieved.

There were no worker or loading dock area contaminations. Radiation Protection surveyed the remaining lead sheets, and several were found to have fixed

4 REPLY TO A NOTICE OF VIOLATION - NRC INSPECTION REPORT 97002 4 contamination. All lead sheets exhibited less than 1000 counts per minute

, -(cpm)'per 100 centimeters squared (cm2) removable contamination except for one from which a hot particle of 4.2 microcuries Cobalt-60 was retrieved.

The stack of lead had been topped by a large, new lead sheet, such that the lower sheets of previously used lead were well protected from the environment.

The depth of lead also provided complete shielding for beta and significant shielding (at least a tenth-value layer) for gamma due to the contamination.

The contamination levels found did not constitute an external wholebody dose hazard.

Following the survey, the lead was placed in a temporary storage / decontamination area in the RCA for decontamination.

The RP Manager reviewed the requirements of Plant Administrative Procedure 5.11, Radioactive Material Control, specifically the survey method for determining the " clean" status of materials which can be released for unrestricted use, with the RP department personnel.

III. The corrective steos that will be taken to avoid recurrence.

1. RM-56, Radiological Clearance of Materials for Offsite Removal, will be revised to be consistent with Administrative Procedure 5.11. This revision will also establish a clearly defined survey area for routine surveys of equipment and materials leaving the Radwaste Facility.

THIS REVISION WILL BE COMPLETE BY SEPTENBER 1, 1997 IV. The date when the facility will be in full compliance.

The facility is currently in full compliance.

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REPLY TO A NOTICE OF VIOLATION - NRC INSPECTION REPORT 97002 5 l l

l Violation 97002-02 1

Technical Specification 3.7(c) requires that "Each reactor shutdown for refueling, but in no case at intervals greater than two years, all l electrical and accessible piping penetration nipple welds be visually \

examined for evidence of corrosion, cracking or deterioration."

Contrary to the above, on January 15, 1997, the Ifcensee discovered that two containment electrical penetrations (H-115A and H-1158), previously  ;

considered inaccessible, were in fact available for inspection but had '

not been inspected since installation in March 1985.

. Consumers Energy Company agrees with the violation as stated.

Pursuant to the provisions of 10 CFR 2.201, a reference to previously docketed ,

material is applicable in this matter, On January 31, 1997, Licensee Etent '

Report (LER) 97001, Test Not Performed in Accordance with Appendix J to 10 CFR Part 50, was submitted. Corrective action associated with the LER is complete. l The Big Rock Point staff believes that this correspondence adequately l addresses the required respon's e.

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j REPLY TO A NOTICE OF VIOLATION - NRC INSPECTION REPORT 97002 6

. i Violation 97002-03 4

10 CFR 50.59 requires, in part, that changes in the facility, as described in SAR (FHSR at Big Rock Point), shall not be made without prior NRC approval unless it is determined that the change does not

represent an unreviewed safety question (USQ), and a written safety.
evaluatton documents the bases for this determinatfon.

4 Contrary to the above, on February 7, 1985, the ifcensee failed to

perform a written safety evaluation to document the basis for i determining that a USQ did not exist when it was determined that i

containment post accident temperatures would exceed those identified in the FHSR.

l j Consumers Energy Company requests that the NRC review the following l- supplemental information that has been retrieved since the time-frame of the inspection to determine if a violation of NRC requirements has occurred.

Basis for further consideration

! Overview i

On October 23, 1996, the Big Rock Point staff concluded that inconsistencies existed in a computer code (CONTEMPT-LT/28) that is used to calculate the i

temperature in the containment building during a primary coolant system leak.

This issue surfaced when the Probabilistic Risk Assessment staff was

! validating a personal computer (PC)-based version of CONTEMPT-LT/28 code.

, The Nuclear Regulatory Commission was notified of the discovery and the

! facility commenced a normal and orderly shutdown at 1135 that same day. Cold j shutdown was reached on October 24, 1996, at 1123.

! The initial evaluation concluded that a correction made to the computer code 1 in August 1983 was incorrect, and that the discrepancy was identified in

! September 1990 through an evaluation by the vendor of the code. The vendor i then notified'all licensed users of their program. Both of Consumers Energy

, Nuclear Power facilities, Big Rock Point and Palisades, used the computer

! code; but only Palisades was identified as the registered user. The computer l

code users at Palisades received the notice of the discrepancy, but discrepancies were not communicated to the users at Big Rock Point.

i j -Big Rock Point has re-evaluated the containment response to hypothetical

! ' design base accidents using the CONTEMPT-LT/28 PC computer code and  !

) demonstrated that when the 1980 and 1982 analyses are compared to the 1996

! CONTEMPT analysis, only slight differences exist between the traces. These

! differences occurred as a result of changing platforms and compilers. The

, input data remained the same. These differences were not significant;

therefore, the 1996 version of CONTEMPT-LT/28 is predicting results the same

{ as that of the 1980 analysis of record.

i i The initial evaluation assumed that a sensitivity study performed in 1985 to

evaluate the effect of higher initial temperatures in containment was the code l' of record. A steam line break sensitivity analysis using conservative input '

changes had been performed to verify that the highest containment temperature alarm setpoint (120'F) would not result in a condition that would have detrimental effects to containment and components within. Previous Consumers

[ Energy analyses had assumed an initial containment temperature of 100*F. The

REPLY TO A NOTICE OF VIOLATION - NRC INSPECTION REPORT 97002 7 evaluation concluded that temperatures up to and including 120*F did not

, violate the plant's design basis. In 1996, this analysis was referenced as the code of record, and it should not have been. The Big Rock Point staff learned through further investigation that the 1980 Code is the analysis of record. ,

Since this activity was a sensitivity analysis only, and there were no changes in plant equipment or procedures, a safety evaluation was not written. '

Discussion This issue surfaced when the Probabilistic Risk Assessment staff was validating a personal computer (PC)-based version of CONTEMPT-LT/28 code. 1 Previous to this, the code existed as a mainframe application only. Converting l to the PC version would provide Big Rock Point the ability to evaluate containment response to potential Loss of Cooling Accidents (LOCAs) and steam line breaks at the site instead of through the General Office. Validation would require reviews of a 50 lb/sec, 0.05 ft2, and a 0.63 ft2 steam line )

break.

Initial validation of the Big Rock Point " VENT" model (which included containment ventilation) of CONTEMPT-LT/28 resulted in excellent agreement with steam line break analyses performed in 1985 (A-NL-84-58, " Big Rock Point Containment Response Analysis for Main Steam Line Breaks", 8/29/85) with regard to the 50 lb/sec steam line break assuming no containment sprays in operation. Codes were compared using the same input data. Following this review, a 0.05 ft2 steam line break was performed. Again, the same input data was used by both codes. This time, the results disclosed a discrepancy between the codes. The difference was due to the method of handling spray energy and spray mass within the superheated atmosphere region (i.e., the amount of ,

containment spray droplet vaporization). In the 1985 analysis, 100% spray I efficiency was assumed. The present analysis assumes 73% spray efficiency, and different spray and mass equations.

A detailed analysis of the 1985 model was not possible because the source code was compiled on a mainframe computer. Therefore, the PC version of the CONTEMPT-LT/28 computer code was modified by an independent assessor to evaluate the discrepancy. The independent assessment was able to approximate the encountered code discrepancy by re-evaluating both the spray mass and energy balance equations, and by modifying the spray fraction efficiency. This resulted in similar curve profiles and a peak temperature difference of about 10*F. These results were deemed to be acceptable t,ecause the peak temperature and curve profile remained below the UFHSR EQ Profile (a.k.a. Figure 3-4) envelope.

Following the 0.05 ft2 steam line break assessment, the 0.63 ft2 steam line break was reviewed. The same input data, data from the 1985 study, was used by both codes. The 1985 code resulted in a faster temperature decay trace and the CONTEMPT-LT/28 PC version calculated a trace with slightly hotter temperatures. This discrepancy was again the result of modifying both the spray mass and energy balance equations and the spray fraction efficiency.

These results prompted an analysis of past CONTEMPT data sets, using the 0.63 ft2 steam line break.

1980, 1982, and 1996 Large Steam Line Break Analysis Consumers Energy Company recognized in 1980, while performing steam line break analysis, that the containment design temperature could be exceeded. A Technical Specification Change Request was submitted on December 5,1980 to address a new calculated peak temperature of 253*F, and was later approved by

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j REPLY TO A NOTICE OF. VIOLATION - NRC INSPECTION REPORT 97002 8  ;

I the Commission as Amendment 37 by letter forwarded January 13, 1981. This 1 i , analysis used CONTEMPT-0T/26 developed in 1975 by Aerojet Nuclear Co., which i

was converted to a Control Data Corporation (CDC) mainframe by Energy j Incorporated (EI) in January of 1977, similar to CONTEMPT-LT/28.

! In 1982, Lawrence Livermore National Laboratory (LLNL) prepared a Technical j' Evaluation Report for the Nuclear Regulatory Commission during the evaluation i of Big Rock Point Systematic Evaluation Report (SEP) Topics VI-2.D, Mass and

Energy Release for Postulated Pipe Breaks Inside Containment, and VI-3, 1

Containment Pressure and Heat Removal Capability. This report used CONTEMPT-i LT/28 which included the 0.63 ft2 steam line break This report was used to '

l close the two SEP topics discussed above.

When the 1980 and 1982 analyses were compared to the 1996 CONTEMPT analysis, i slight differences existed between the traces. These differences occurred as a l 1

result of changing platforms and compilers. The input data remained the same.  !

However, these differences were not significant, therefore the 1996 version of

  • CONTEMPT-LT/28 is predicting results comparable to the 1980 analysis of
record.

4 Electrical Equipment Qualification I In 1983, the NRC issued the Safety Evaluation Report (SER) for the

! Environmental Qualification of Safety-Related Electrical Equipment. One of the i major qualification deficiencies that had been identified and required special attention resided in the containment pressure / temperature profiles defining

acceptable post-accident environmental envelopes for use in equipment

! qualification at Big Rock Point. The NRC was concerned that the current,1975 l EEQ profile did not take into account the most limiting line breaks (recirculation line break and the 0.63 ft2 main steam line break) that had i been discussed during the approval of Amendment 37. In a 1984 Consumers Energy j' memo, a proposal to provide a composite envelope of the 1975 envelope and

incorporation of the breaks discussed above was made, The revised envelope j took into consideration the Consumers Energy Company's calculation of peak containment temperature (253'F), and the NRC's calculation from the-SER l (251*F). Following a discussion with the NRC, an envelope of 260*F was chosen.
The EEQ profile was updated and placed in the EEQ File on November 7, 1984.
Equipment in containment required to operate after a large steam line break l was qualified to the revised temperature profile.

i t 1980 and 1985 Comparisons

Input data changes occurred between the 1980 analysis of record, and the 1985 sensitivity analysis. The changes, containment free volume, initial i containment atmosphere, containment shell not credited as a heat sink, f structural steel surface area, lead shield plug thickness and surface area, i aluminum structures surface area, and sheet metal cabinet thickness, collectively resulted in a slightly higher temperature profile (~ 13*F).

! The 1985 analysis is the only analysis that has been discussed that slightly j transcends the UFHSR EQ Profile (a.k.a. Figure 3-4) envelope. On February 7, i 1985, A-BRP-85-01 was initiated. The Action Item Record requested that a steam

} line break sensitivity analysis be performed to verify that the highest j containment temperature alarm setpoint (120*F) would not result in a condition

that would have detrimental effects to containment and components within.

! Previous Consumers Energy analyses had assumed an initial containment i

temperature of 100'F. The evaluation concluded that temperatures up to and t including 120*F did not violate the plant's design basis. In 1996, this

analysis was referenced as the code of record, and it should not have been.

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REPLY TO A NOTICE OF VIOLATION - NRC INSPECTION REPORT 97002 9 The Big Rock Point staff learned through further investigation that the 1980

, -Code is the analysis of record. Further, the Big Rock Point staff believes that the EQ 260*F envelope boundary, revised November 7, 1984, was not revised to 266*F because of the conservative input changes, and that the current boundary was exceeded for just a short time period thereby precluding any significant component heatup demonstrated by detailed thermal lag results  ;

assuming containment temperature loads in excess of 266*F. Since this activity  !

was a sensitivity analysis only, and there were no changes in plant equipment or procedures, a safety evaluation was not written. 1 4

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REPLY TO A NOTICE OF VIOLATION - NRC INSPECTION REPORT 97002 10 Violation 97002-04

^ 10 CFR 50.49 requires, in part, that the licensee establish an equipment qualification program which must include and be based on the time-dependent temperature at the location of the equipment following the most severe design basis accident for which this equipment is required to remain functional.

10 CFR 50.71(e) requires, in part, that FSAR (FHSR at Big Rock Point) updates shall contain all changes necessary to reflect analyses prepared pursuant to NRC requirements.

Contrary to the above, on December 22, 1989 the licensee submitted a FHSR update which failed to reflect the results of analysis _on time-dependent temperatures inside containment following design basis accidents, which had not been prepared pursuant to 10 CFR 50.49.

This is a Severity Level IV violation (Supplement I).

Consumers Energy Company requests that the NRC review the following supplemental information that has been retrieved since the time-frame of the

, inspection to determine if a violation of NRC requirements has occurred.

i Basis for further consideration During the Integrated Plant Safety Assessment; Systematic Evaluation Program (SEP), 10 CFR 50.71(e)(3)(ii) was issued. The SEP was initiated in February, 1977 by the NRC to review the designs of older operating nuclear reactor plants to reconfirm and document their safety. This required in part, that licensees of nuclear power plants subjected to the SEP, file an updated Final Hazards Summary Report FHSR within 24 months following notification from the NRC that the SEP has been completed. In evaluating the benefit of this action as part of the Integrated Plant Safety Assessment for Big Rock Point (NUREG-0828 - Final Report; May 1984, section 5.3.25), Consumers Energy Company proposed to evaluate a method of indexing existing documents to provide a workable substitute versus expending the resources necessary for a complete FHSR update. In a letter dated December 4,1984, the NRC following review of the Big Rock Point submittal, supported Consumers Power Company's alternative to an updated FHSR. However, just prior to the FHSR cross index submittal in August of 1986, the NRC identified concerns with the developed cross index system. As a result, the Staff requested that Consumers Energy Company complete an FHSR update and that a scheduler exemption to allow time to perform this effort was warranted.

In a letter dated December 3, 1986, Consumers Energy Company committed to perform an FHSR rewrite. The goal of this effort was to provide a document that provided a current snapshot of the Big Rock Point design and could be used by the plant staff to perform 10 CFR 50.59 evaluations. On December 22, 1989, the Updated FHSR was submittad to the NRC. That is when data from the EEQ Files (i.e., Containment Temperature vs. Time Profile; Figure 3-4) was initially included in the r riSR.

In summary, the 1985 sensitivity analysis addressed in Violation 97002-03 did not require a change to the UFHSR, and the 1984 changes required for EEQ as committed to the NRC were made as part of the 1989 UFHSR submittal.

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