U-603212, Forwards Response to NRC 990315 Telcon Re Proposed Amend of License NPF-62,implementing Feedwater Leakage Control Sys Mode of RHR Sys

Forwards Response to NRC 990315 Telcon Re Proposed Amend of License NPF-62,implementing Feedwater Leakage Control Sys Mode of RHR Sys

ML20212J313
Person / Time
Site:	Clinton
Issue date:	06/24/1999
From:	Mcelwain J ILLINOIS POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
U-603212, NUDOCS 9906290187
Download: ML20212J313 (10)

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l John P McEmain

.."# _e.' .a P6WER u-60m2 An Illinova Company June 24, 1999 Docket No. 50-461 10CFR50.90 Document Control Desk Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

Clinton Power Station AdditionalInformation Regarding Proposed Amendment of Facility Operating License No. NPF-62 (LS-97-006)

Dear Madam or Sir; By Illinois Power (IP) letter U-603032, dated October 23,1998, (and as supplemented by IP letter U-603160, dated Febmary 22,1999) IP requested amendment of the Clinton Power Station (CPS) Operating License (License No.

NPF-62) pursuant to 10CFR50.90. The application, which is currently under review by the NRC staff, consists of a proposed change to implement a feedwater leakage control system (FWLCS) mode of the residual heat removal (RHR) system. If approved, the amendment will change the periodic leakage testing requirement for the primary containment feedwater penetration isolation valves such that a water leakage test would I

be performed in lieu of the presently required air leakage test.

In support of the amendment request, a teleconference between the NRC staff /

and representatives from IP was held on March 15,1999. During this teleconference the NRC staff requested additional information with respect to the time delay assumed in the radiological dose analysis for initiation of the high radiation mode of the main control room ventilation system due to single-failure considerations associated with that system. Attachment 2 of this letter contains the requested additional information.

Attachment 2 of this letter also provides an update / revision of the summary table of radiologicad effects associated with the subject license amendment, which was '. s included in IP's request dated October 23,1998, and as updated in IP's letter dated February 22,1999. As explained in Attaclunent 2, this latest revision / update to the summary table of radiological effects has become necessary as a result of corrective action taken to address a documentation discrepancy (previously unrelated to the 9906290187 990624 PDR ADOCK 05000461 P PDR

i U403212 Page 2 requested changes to the CPS Operating License) wherein the actual net free volumes of the primary containment and drywell have been determined to be slightly smaller than the corresponding values used for these parameters in the original plant analyses. (This non-conforming condition is the subject of CPS Licensee Event Report 1999-006.) Operability of affected plant systems, structures, and components was resolved by performing bounding analyses to show that the impacts of the revised prinwy containment and drywell volumes are acceptable and that the primary containment and drywell will still operate within their design parameters during all analyzed events. A summary of this l' operability evaluation is included in Attachment 2.

The net free volume discrepancy issue impacts post-accident dose consequences because the reduced primary containment and drywell net free volume increases the calculated post-accident radionuclide concentration in the containment and drywell atmosphere thereby increasing the radiological dose consequences associated with leakage l

from the main steam isolation valves (MSIVs). Recent analyses have demonstrated that l with the use of administrative controls the post-accident main control room and offsite doses remain within the current CPS licensing basis values. The calculated doses, l however, without the use of administrative controls are slightly higher than those in the current CPS licensing basis and thus require review and approval by the h%C. Since the FWLCS amendment request also impacts the calculated post-accident doses, these two

! activities are being combined to promote an efficient and effective review process.

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As noted previously, Attachment 2 contains a revision to the summary of radiological effects (Table 3) calculated for the associated FWLCS license amendment since the previous calculations performed in support of this amendment request did not l include consideration of the reduced primary containment and drywell net free volumes.

Consequently, the revised sununary of radiological effects updates and supersedes the l . similar tables that were provided in IP's October 23,1998, and February 22,1999 letters.

l Sincerely yours,  ;

1 L ohn P. McEl ain ChiefNuclear Officer JFK/krk Attachments cc: NRC Clinton Licensing Project Manager NRC Resident Office, V-690 {

- Regional Administrator, Region III, USNRC lilinois Department of Nuclear Safety i

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l Attachment 1 I

to U-603212 l Page1ofI l

l John P. McElwain, being first duly sworn, deposes and says: That he is ChiefNuclear Officer for Clinton Power Station; that this application for amendment of Facility l . Operating License NPF-62 has been prepared under his supervision and direction; that he knows the contents thereof, and that to the best of his knowledge and belief said letter and the facts contr.ined therein are tme and correct.

l l

Date: This 2 day of June 1999.

l Signed: M 7M -[- '

, [ ~ John P/Mestwain STATE OF ILLINOIS l SS.

3d* COUNTY j Subscribed and sworn to before me this W day ofJune 1999.

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• OFFICIAL SEAL

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l Attachment 2 to U-603212 Page 1 of 7 Response to Request for Additional Information During a March 15,1999 teleconference between the NRC staff and representatives from IP, the NRC staff requested additional information with respect to the time delay assumed for initiation of the high radiation mode of the main control room ventilation system as assumed to occur for restoring the VC system following postulated single failures.

As described in IP's February 22,1999 letter (which provided additional information and )

supplemented IP's initial request dated October 23,1998) corrective action was necessary to )

address a discrepancy between the design of the main control room ventilation (VC) system l and the assumptions used in the original post-accident radiological dose analysis. Specifka.ij, the calculation of the design basis accident (DBA) dose consequences to the main control room operators assumed that the VC system would be running and supplying filtered, pressurized air to the control room envelope immediately following and throughout the duration of the accident. However, a review of the VC system was performed and subsequently disclosed that for postulated single failures (e.g., loss of an emergency diesel generator, failure of a fan or a damper, etc.) manual operator action would be required to restore at least one train of the VC system in the high radiation mode of operation.

The r&diological dose calculations performed in support of the FWLCS license amendment request were revised to resolve this discrepancy by including consideration of the most limiting single-active failures of the VC system. For dose analysis purposes, it was assumed (based on Standard Review Plan Section 6.4.III.3.d.3.ii) that the high radiation mode of VC operation is unavailable for the first 20 minutes of the accident. The analyses subsequently assumes that the control room HVAC air flows are returned to the design basis values 20 minutes after initiation of the accident. This 20-minute time delay includes time for the VC failure to be revealed (in response to an automatic accident signal), for operators to take action in response to the failure, and the time required for the system to achieve design conditions. This time delay is commensurate with delay times assumed for operator actions in other analyses and as recognized in the Standard Review Plan.

All of the postulated single active failures of the VC system were bounded by eight general cases. It was found that one case resulted in the most limiting thyroic. dose, while a different case generated the most limiting y-whole body and p-skin doses. Thus, the most limiting  ;

doses from both of these cases are entered in the summary table of radiological effects (Table 3). The main control room data used for calculating the doses associated with the two most-limiting single failure cases for the first 20 minutes after the postulated accident are shown in Table 1. Table 1 also includes a column of data used to calculate the doses after the first 20 minutes.

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Attachment 2 to U-603212 Page 2 of 7 Reduced Primary Containment and Drywell Net Fue Volumes As discussed in the cover letter, a discrepancy pertaining to the net free volume of the primary containment and drywell has recently been identified (LER 1999-006). Specifically, investigations have revealed that the actual primary containment and drywell net free volumes are smaller than the values that were used as input for the original plant analyses. The primary containment volume used in the original analyses was 1,550,800 cubic feet, while the actual value is approximately 1,512,300 cubic feet. Similarly, the drywell volume used in the original analyses was 246,500 cubic feet, while the actual volume is approximately 241,700 cubic feet.

Collectively, this represents a volume reduction of about two and one-half percent.

These volume discrepancies have a puential impact on the evaluations of the primary containment and drywell responses during postulated accident conditions described in the CPS Updated Safety Analysis Report (USAR) Section 6.2.1, " Containment Functional Design," as well as other USAR sections. As a result, bounding analyses were performed to determine the impact of the smaller volumes on the accident analyses.

These bounding analyses, including consideration of the impact on various dynamic loads, determined that although there were severalinstances where specific calculated parameters (e.g., drywell pressure and drywell negative pressure) slightly exceeded the values currently listed in the USAR, all of the affected design parameters remained below their corresponding design limits. However, one of these design parameters (containment pressure due to a small break loss-of-coolant accident) required imposing an administrative compensatory action (reducing the allowable drywell bypass leakage area from 1.18 square feet to 1.0 square feet) in order to meet the primary containment pressure design limit (as described in LER 1999-006).

Additional investigations were performed to determine the impact on other supporting calculations created by this volume discrepancy. The only significant impact that was identified pertained to the post-accident radiological dose analyses. Since the post-accident radioactive source term (fission product inventory) is a constant value, the concentration of radioactivity in the containment atmosphere assumed in the dose analyses will increase due to the decrease in the primary containment and drywell volumes. This resulting higher concentration of radioactivity increases the post-accident doses that are affected by main steam isolation valve (MSIV) leakage. With no compensatory actions, calculations show that some of the analyzed doses increase slightly above their current licensing basis values as a result of the smaller primary containment and drywell volumes. To compensate for this increase, administrative controls were invoked to reduce the MSIV leakage limit through each main steam line from its Technical Specification Surveillance Requirement (TS SR 3.6.1.3.9) value of 28 standard cubic feet per hour (scfh) to 27 scfh. This is an acceptable approach since a review of the most recent test results showed that the total leak rate for the MSIVs is well below the new administrative limit (about 40 percent of the allowable total). With these administrative controls in place, all of the post-accident doses remain at or below their previously analyzed values.

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Attachment 2 to U-603212 Page 3 of 7 f

l Table 2 centains the revised data associated with the primary containment and drywell net free l

volume issue. These data were used to calculate the revised doses listed in the summary table of radiological effects (Table 3) associated with the subject license amendment request. After approval of this license amendment request (which will establish new licensing basis limits with respect to the DBA dose analysis), the MSIV administrative limit of 27 scfh will no j

longer be required and the leakage limit will be retumed to the Technical Specification value l

of 28 scfh.

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The proposed licensing basis doses listed in Table 3 are based on doses obtained from DBA analyses for the revised leakage pathways introduced by the FWLCS. The proposed dose values also reflect consideration of the most limiting single-active failure of the VC system and account for the actual net free volume of the primary containment and drywell. There are also I several conservatisms taken in the analysis. For example, a 300 second drawdown time for the secondary containment envelope is assumed in the analysis in lieu of the licensing basis requirement of s 188 seconds. An additional conservatism taken is that the calculated doses were increased by approximately ten percent (rounding up to the next u hole rem for the  ;

thyroid and p-skin doses, and rounding up to the next one half rem for the y-whole body l doses). The conservatisms used in the analyses supporting this license amendment, including  !

the approximate ten-percent margin, are not a requirement for radiological analysis. Rather, j

they are included in this license amendment request to bound such considerations as  ;

instrument uncertainties, and provide operational margin for use in any subsequent minor {

radiological analyses that would not warrant the processing of a license amendment. l Operability Considerations I

As noted previously (and in the cover letter), identification that the drywell and primary containment volumes are slightly smaller than what was previously assumed in the plant safety analyscs constitutes identification of a nonconforming condition. With regard to determining operability for the affected systems, structures, and components, Generic Letter 91-18,  ;

Revision 1, "Information To Licensees Regarding NRC Inspection Manual Section On l Resolution Of Degraded And Nonconforming Conditions, Change to Current Licensing )

Basis," dated October 8,1997, includes the following discussion regarding changing the licensing basis to accept a nonconforming or degraded ccndition:

[One) situation (to consider) is a final resolution in which the licensee proposes to change the current licensing basis to accept [an] as-found nonconforming condition. In 4 this case, the 10 CFR 50.59 cvaluation is of the change from the SAR-described condition to the existing condition in which the licensee plans to remain (i.e., the )

licensee will exit the corrective action process by revising its licensing basis to document acceptance of the condition). If the 10 CFR 50.59 evaluation concludes that a change to the TS or a USQ is involved, a license amendment must be requested, and the corrective action process is not complete until the approval is received, or I other resolution occurs. In order to resolve the degraded or nonconforming condition j without restoring the affected equipment to its original design, a licensee may need to I obtain an exemption from 10 CFR Part 50 in accordance with 10 CFR 50.12, or relief I from a design code in accordance with 10 CFR 50.55a. The use of 10 CFR 50.59, )

50.12, or 50.55a in fulfillment of Appendix B corrective action requirements does not I

r Attachment 2 to U-603212 Page 4 of 7 relieve the licensee of the responsibility to determine the root cause, to examine other affected systems, or to report the original condition, as appropriate.

In both of these situations, the need to obain NRC approval for a change (e.g.,

because it involves a USQ) does not affect the licensee's authority to operate tim plant.

The licensee may make mode changes, restart from outages, etc., provided that necessary equipment.is operable and the degraded condition is not in conflict with the TS or the license. The basis for this position was previously discussed in Section 4.5.1.

Operability of plant systems, structures, and components affected by the containment net free volume issue was resolved by performing bounding analyses to show that the impacts of the revised primary containment and drywell volumes are acceptable and that the drywell and containment will still operate within their design parameters during all analyzed events.

Similarly, a reanalysis of the dose consequences from a postulated design-basis loss-of-coolant accident was performed. This reanalysis included consideration of the smaller-than-previously analyzed primary containment and drywell net free volume. The results demonstrated that the doses received by the main control room operators were oclow the limits of General Design Criterion 19 of Appendix A to 10CFR50 and that offsite doses remain below the limits of j

10CFR100.11. Thus, the primary containment and drywell were determined to be operable on the basis that they are performing per design, and the associated radiological consequences remain within the limits of the regula*ory requirements.

Based on the above, Illinois Power has determined that the nonconforming / degraded condition is not in conflict with the Technical Specifications (TS) or the license. Since the TS are satisfied and required equipment is ' operable, and since IP is correcting the degraded or nonconforming condition in a timely manner (i.e., requesting NRC approval for the change via this amendment request), continued plant operation does not pose an undue risk to public health and safety.

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6 Attachment 2 to U403212  !

Page 5 of 7 Table 1 Control Room HVAC Data Used to Generate the Licensing Basis Doses i

Postulated Single Failure Postulated Single Failure Nominal Base of the VC System of the VC System Value (applies Resulting in the Most Resulting in the Most after 20 minutes)

Limiting Thyroid Dose Limiting y-Whole Body and -Skin Doses Main Control Room 3.24E+05 3.24E+05 3.24E+05 Free Volume (ff)

Make-up Air Flow Rate 0 3920 3300 f (cfm)  !

I Recirculation Air Flow 0 54900 54900 Rate (cfm) .

5 Make-up Air Filter 0.99Y 0.99 0.99

• Fractional Efficiency for Iodines Recirculation Air Filter 0,70Y- 0.70 0.70 Fractional Efficierry for Iodines Unfiltered Inleakage 1510 10 10 1 Rates (cfm)

Filtered Inleakage Rates 0 650 650 (cfm) y1 = Since make-up and recirculation flows are zero for this case, the effective fil[er efficiency is also zero.

T = For this case, the overall effective makeup air filtn efficiency is equal to 0.9488 and is derived by the following relationship: l (1-EE) = [(1 - EM)(FM) + (t - EFI)(FFI)] / (FM + FFI)

. where: EE = effective filter efficiency EM = design basis make-up filter efficiency (0.99)

EFI = design basis recirculation filter efficiency (0.70)

FM = makeup air flow rate (cfm)

FFI = filtered inleakage air flow rate (cfm) c = For this case, the overall effective makeup air filter efficiency is equal to 0.9423 and is also derived by the relationship given above.

e Attaciunent 2 to U-603212 Page 6 of 7 Table 2 Primary Containment Data Used to Generate the Licensing Basis Doses Leak Rate (%/ day) 0.65 Containment Free Volume (R') 1.512E+06 Drywell Free Volume (R') 241,000 Projected Area Normal to Wind (m') 1460 MSIV Leak Rate (SCFH perline) 28 Suppression Pool Vent Area (A') 420 ElementalIodine Fraction 0.91 Particulate Iodine Fiaction 0.05 Organic lodine Fraction 0.04 SGTS Fractional efficiency for Iodines 0.99 (aRer drawdown)

Suppression Pool Water Volume (R') 146,400 Secondary Containment Bypass Fraction 8%

Drawdown Time (sec) 300(0

)

(1) 300 seconds was used for conservative analyses purposes. The CPS licensing basis drawdown time remains at s 188 seconds.

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a Attachment 2 to U-603212 Page 7 of 7 Table 3 Summary of Radiological Effects (Rem) 30-Day Current CPS Effect of using FWLCS dose Proposed new 10CFR50 MCR- Licensing Basis -POSTDBA, calculation CPS Licensing Appendix A Dose (as ICRP 30, *

• results *

• Basis Dose + c GDC 19 Limit redected in the and suppression USAR) (DBA-LOCA) (SRP 6.4.II 6) pool scrubbing Thyroid 4.934 4.3 (27) A 21.87 25 30 7-Whole 2.0 1.945 3 6 5 2.425 3.0 5 Body

-Skin 14.3 13.96 17.63 20 30 2-Hour Current CPS Effect of usmg FWLCS dose Proposed new 10CFR100.11 EAB Licensing Basis POSTDBA, calculation CPS Licensing Limit Dose (as ICRP 30,

• results

• Basis Dose

• renected in the and suppression USAR) (DBA-LOCA) pool scrubbing Thyroid 163 19.51 204.4 225 300 y-Whole 4.4 3.742 9.948 11 25 Body 30-Day Current CPS Effect ofusing FWLCS dose Proposed new 10CFR100.11 LPZ Licensing Basis POSTDBA, calculation CPS Licensing Limit Dose (as ICRP 30,

• results

• Basis Dose

• reflected in the and suppression USAR) (DBA-LOCA) pool scrubbing Thyroid 156 18.14 77.56 86 300 y-Whole 1.7 1.556 3.022 3.5 25 Body The 27 rem value includes 650 cfm of filtered duct leakage. The 4.3 rem value does not include duct leakage.

For calculating MCR y-whole body doses, Reg. Guide 1.109 dose conversion factors were used in conjunction wit correction factor from Murphy-Campe. This value also includes dose received from external shine sources.

• ICRP 30 dose conversion factors were used for thyroid doses. Reg. Guide 1.109 dos -

body and p-skin doses.

• MCR doses include consideration of the most limiting single failure of the main control room ve also include 650 cfm of filtered duct leakage.

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