ML12124A283

Response to Request for Additional Information: License Amendment Request 244, Proposed Revision to Radiological Accident Analysis and Control Room Envelope Habitability Technical Specifications

ML12124A283
Person / Time
Site:	Kewaunee
Issue date:	04/30/2012
From:	Price J Dominion, Dominion Energy Kewaunee
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
12-214, TAC ME7110
Download: ML12124A283 (48)

Text

Dominion Dominion Energy Kewaunee, Inc.

5000 Dominion Boulevard, Glen Allen, VA 23060 April 30, 2012 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Serial No.12-214 LIC/CDS/R6 Docket No. 50-305 License No. DPR-43 DOMINION ENERGY KEWAUNEE, INC.

KEWAUNEE POWER STATION RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION:

LICENSE AMENDMENT REQUEST 244, PROPOSED REVISION TO RADIOLOGICAL ACCIDENT ANALYSIS AND CONTROL ROOM ENVELOPE HABITABILITY TECHNICAL SPECIFICATIONS (TAC NO. ME7110)

By application dated August 30, 2011 (Reference 1), Dominion Energy Kewaunee, Inc.

(DEK), requested an amendment to Facility Operating License Number DPR-43 for Kewaunee Power Station (KPS). This proposed amendment (LAR 244) would revise the KPS Operating License by modifying the Technical Specifications (TS) and the current licensing basis (CLB) to incorporate changes to the current radiological accident analysis (RAA) of record. This amendment would also fulfill a commitment made to the NRC in response to Generic Letter 2003-01, "Control Room Habitability" (References 1 and 2) to submit proposed changes to the KPS TS based on the final approved version of TSTF-448, "Control Room Habitability."

Subsequently, on March 2, 2012 the Nuclear Regulatory Commission (NRC) staff transmitted a request for additional information (RAI) regarding the proposed amendment (Reference 3).

The RAI questions and associated DEK responses are provided in Attachment 1 to this letter.

Serial No.12-214 LAR 244 RAI Response Page 2 of 3 If you have any questions or require additional information, please contact Mr. Craig Sly at 804-273-2784.

Sincerely, Price V"

"eesident

- Nuclear Engineering STATE OF CONNECTICUT COUNTY OF NEW LONDON The foregoing document was acknowledged before me, in and for the County and State aforesaid, today by J. Alan Price, who is Vice President - Nuclear Engineering, of Dominion Energy Kewaunee, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this.'

day of /191XZ

,2012.

My Commission Expires: f6'*?iUA,";

2? Z01/6

-N a

Notary Pub

Attachment:

1. NRC Request for Additional Information Questions and Dominion Energy Kewaunee Responses

Enclosure:

1. Copy of Letter from K. E. Perkins (NRC) to D. C. Hintz (WPS) dated February 16, 1988 Commitments made in this letter: None

I Serial No.12-214 LAR 244 RAI Response Page 3 of 3

References:

1. Letter from J. A.. Price (DEK) to Document Control Desk (NRC), "License Amendment Request 244, Proposed Revision to Radiological Accident Analysis and Control Room Envelope Habitability Technical Specifications," dated August 30, 2011. [ADAMS Accession No. ML11252A521]

2. Letter from Craig W. Lambert (NMC) to Document Control Desk (NRC), "Generic Letter 2003-01; Control Room Habitability - Supplemental Response," dated April 1, 2005. [ADAMS Accession No. ML050970303]

3. E-mail from Karl D. Feintuch (NRC) to Craig D. Sly (DEK), "ME7110 Kewaunee Amendment Request Re: Chi-over-Q - AADB Request for Additional Information (RAI)," dated March 2, 2012. [ADAMS Accession No. ML12066A008]

cc:

Regional Administrator, Region III U. S. Nuclear Regulatory Commission 2443 Warrenville Road Suite 210 Lisle, IL 60532-4352 Mr. K. D. Feintuch Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08-H4A 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Kewaunee Power Station Public Service Commission of Wisconsin Electric Division P.O. Box 7854 Madison, WI 53707

Serial No.12-214 ATTACHMENT 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION:

LICENSE AMENDMENT REQUEST 244, PROPOSED REVISION TO RADIOLOGICAL ACCIDENT ANALYSIS AND CONTROL ROOM ENVELOPE HABITABILITY TECHNICAL SPECIFICATIONS NRC REQUEST FOR ADDITIONAL INFORMATION QUESTIONS AND DOMINION ENERGY KEWAUNEE RESPONSES KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Serial No.12-214 Page 1 of 35 NRC REQUEST FOR ADDITIONAL INFORMATION QUESTIONS AND DOMINION ENERGY KEWAUNEE RESPONSES On March 2, 2012 the Nuclear Regulatory Commission (NRC) staff transmitted a request for additional information (RAI) (Reference 3) regarding Dominion. Energy Kewaunee, Inc. (DEK) proposed amendment LAR 244 (Reference 1).

The RAI questions and associated DEK responses are provided below.

1.

NRC Question 1 (ME7110-RAII-AADB-Blu.m-001-2012-03-02), of the proposed license amendment request (LAR) (Adams Package No. ML11252A521), page 91 states:

The dose conversion factors used to calculate the TEDE doses and DE 1-131 for the Steam Generator Tube Rupture accident were taken from Table 3.2-3 for the isotopes required by Regulatory Guide 1.183 for the SG TR analysis.

Table 3.2-3 provides the effective dose equivalent (EDE) and committed effective dose equivalent (CEDE) dose conversion factors for iodine. Therefore, according to the statement above, the total effective dose equivalent (TEDE) is calculated with either the EDE or CEDE dose conversion factors. page 32 of the LAR states:

The dose conversion factors (DCFs) used to determine dose from iodine are from Federal Guidance Report No. 11 (FGR-11), Table 2.1 committed effective dose equivalent (CEDE) and the calculation of the Dose Equivalent 1-131 from proposed technical specification surveillance are from FGR-11 Table 2.1 Thyroid Committed Dose Equivalent (CDE).

The two cited texts appear to conflict. Please clarify which dose conversion factors are used for each design basis (DBA) accident (CDE vs. CEDE). Since the TEDE is defined as the DDE plus the CEDE, please justify use of CDE dose conversion factors.

DEK Response:

In all of the LAR-244 analyses TEDE dose is calculated using the EDE DCFs from FGR-12 and CEDE DCFs from FGR-11 listed in LAR-244 Attachment 4 Table 3.2-3.

The proposed dose equivalent (DE) 1-131 Technical Specification definition using Thyroid CDE DCFs from Table 2.1 of FGR-11 and proposed 0.1 pCi/gm activity limit

Serial No.12-214 Page 2 of 35 were used to determine the relative amounts of iodine isotopes in the Technical Specification reactor coolant system (RCS) source term. The Technical Specification RCS source term was used in the analyses for the LOCA, SGTR, and MSLB.

Although KPS is licensed to 10 CFR 50.67, "Accident source term," the current definition of DE 1-131 requires DE 1-131 to be calculated using ICRP-30 (Reference 4)

DCFs. The amendment request proposed that DE 1-131 be calculated with DCFs from FGR-11, Table 2.1, "Thyroid CDE" (Reference 5). Based on the Reviewer's Note for the definition of DE 1-131, the use of either Thyroid CDE or CEDE DCFs from FGR-1 1, Table 2.1 is specifically allowed by TSTF-490 (Reference 6) for plants licensed to 10 CFR 50.67.

RG 1.183 (Reference 7) requires that the pre-accident and concurrent iodine spikes used in the design basis analysis be based *on the maximum value permitted by Technical Specifications. The proposed 0.1 pCi/gm DE 1-131 inventory was calculated using FGR-1 1 Thyroid CDE DCFs and was used to establish the design basis analysis source term for both the pre-accident and concurrent iodine spikes. The use of FGR-1 1 CEDE DCFs to calculate dose is consistent with the Total Effective Dose Equivalent methodology described in RG 1.183. As shown in the tables below, the use of FGR-11 Thyroid CDE DCFs to perform the Technical Specification surveillance for DE 1-131 will allow a 2% higher total allowable iodine inventory in the RCS than would be attainable using FGR-1 1 CEDE DCFs. The higher source term used in conjunction with CEDE DCFs to calculate TEDE, produce a greater bounding dose consequence that remains below regulatory limits.

FGR-11 0.1 pCi/gm Thyroid CDE DCF 1% FF 1% FF DE 1-131 (Sv/Bq)

(pCi/gm)

DE 1-131 (pCi/gm)

(pCi/gm) 1-131 2.92E-07 2.89E+00 2.89E+00 7.82E-02 1-132 1.74E-09 2.95E+00 1.76E-02 7.97E-02 1-133 4.86E-08 4.31 E+00 7.18E-01 1.17E-01 1-134 2.88E-10 5.97E-01 5.89E-04 1.62E-02 1-135 8.46E-09 2.36E+00 6.85E-02 6.40E-02 total 3.694E+00

Serial No.12-214 Page 3 of 35 FGR-11 0.1 pCi/gm CEDE DCF 1% FF 1% FF DE 1-131 (Sv/Bq)

(pCi/gm)

DE 1-131 (pCi/gm)

(pCi/gm) 1-131 8.89E-09 2.89E+00 2.89E+00 7.64E-02 1-132 1.03E-10 2.95E+00 3.41E-02 7.79E-02 1-133 1.58E-09 4.31E+00 7.66E-01 1.14E-01 1-134 3.55E-11 5.97E-01 2.38E-03 1.58E-02 1-135 3.32E-10 2.36E+00 8.83E-02 6.25E-02 total 3.781 E+00 It is acceptable for the pre-accident and concurrent iodine spike design basis source terms to be based on FGR-1 1 Thyroid CDE DCFs and the doses to be calculated using FGR-1 1 CEDE DCFs because the source term will bound allowable plant operating parameters as defined in the Technical Specifications, as required by RG 1.183, and result in bounding dose consequences.

Serial No.12-214 Page 4 of 35

2.

NRC Question 2 (ME7110-RAII-AADB-Blum-002-2012-03-02), page 45 states:

A reduction in airborne radioactivity in the containment by natural deposition within containment is credited. The model used is described in NUREG/CR-6189, "A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments," (Reference 25) and is incorporated into the RADTRAD-NAI computer code. This model is called the Powers model, set for the 10th percentile., page 43 also discusses credit for removal of aerosol by containment sprays.

Removal of aerosol by sprays and natural deposition are competing processes. Please justify crediting both spray removal and the proposed Powers natural deposition model.

Describe how the Powers natural deposition model accounts for removal due to the spray model used.

If any further credit for a reduction in aerosols is taken for any pathway, please provide a justification for that credit while considering the impact of any other removal mechanism credited. For example, with respect to spray removal and natural deposition ensure that the sprays do not remove aerosols that are also removed in the Powers model.

DEK Response:

The RADTRAD Powers aerosol natural deposition model was used during the entire event, including the period when containment sprays were operating.

However,. the impact of the Powers natural deposition model during the period when containment sprays were operating was negligible. An estimate of the impact of the treatment of aerosols was made using the same model that was used for the LOCA and turning off the RADTRAD Powers aerosol natural deposition model and replacing it with manually-calculated natural deposition decontamination coefficients used only after the containment sprays terminated. The comparison of results, with and without crediting natural deposition during spray removal, shows no difference in resulting dose.

Additional airborne radioactivity removal during the period of competing processes is not apparent. Removal due to Powers natural deposition does become apparent during the remainder of the accident after sprays are terminated.

Serial No.12-214 Page 5 of 35

3.

NRC Question 3 (ME7110-RAII-AADB-Blum-003-2012-03-02), page 46 states that the revised LOCA analysis contains some changes that include:

Replacement of the assumed 1% iodine evolution rate from RWST back-leakage to a conservative DF=100.

Justify this change and describe why it is conservative.

DEK Response:

The current KPS licensing basis for analyzing this release pathway is documented in the Safety Evaluation (SE) related to KPS License Amendment 166 (Reference 8).

Specifically, the leakage water which is at a temperature below 212OF enters the RWST which is significantly below 212'F and enters the tank from the bottom. The evolution of iodine from sump fluids that leak into the RWST is modeled by assuming 1 percent iodine evolution. The current method treats leakage into the RWST the same as any other ESF leakage.

No credit was taken for dilution into existing 'clean' RWST boric acid fluid that remains in the tank nor release into the gaseous 'unfilled' region of the tank.

As specified in the SE for KPS License Amendment 166, "The radiological consequence contribution from this pathway is less significant (less than 2 percent) at the EAB for the postulated LOCA."

As specified in LAR 244 and discussed with the NRC staff during a telephone conversation on October 20, 2011, DEK has chosen to model the release of radioactivity from RWST back-leakage consistent with similar modeling used at other Dominion facilities (References 9 and 10; Millstone Unit 3 TAC Number MC3333; and North Anna TAC Numbers MD3197 and MD3198).

The applied DF values used to model RWST releases for these facilities were 100 and 40, respectively.

Dominion provided previous explanations as to the applicability and appropriateness for this method in a response to an NRC Request for Additional Information (Reference 11) when the method from LAR 244, Attachment 4, Reference 26 ("Iodine Removal from Containment Atmospheres by Boric Acid Spray," Report No. BNP-100, Battelle Memorial Institute, Pacific Northwest Laboratories (PNL), Richland, WA 99352 7/19/70, A. K. Postma, L. F. Coleman and R. K. Hilliard) was first employed to estimate iodine partition coefficients in an RWST for Millstone Unit 3. The emphasis in that response focused on the fact that an RWST should be treated as a closed system for the establishment of achieving equilibrium conditions between the water and air in the tank.

The environment in the RWST does not experience high temperature, radiation, or forced ventilation. It is a static environment that will experience partitioning between the higher iodine concentrations in the water that will drive to achieve equilibrium with the iodine concentration in the air.

Serial No.12-214 Page 6 of 35 The importance of pH in iodine evolution is understood and was considered in the application of iodine release from an RWST.

The RWST environment does not experience high doses which relate to iodine radiolysis discussed in more recent studies performed on iodine evolution and the importance of pH control to hinder the formation of 12, (e.g., NUREG/CR-5950 and NUREG/CR-4697).

It is the formation of hydrogen peroxide (H20 2) during the irradiation of water that primarily reacts with iodide (I ) to form 12. For pools with sufficient iodine concentration that contend with the interaction of iodine species and the products of water radiolysis, the effect of pH has been shown to be a sensitive parameter in iodine evolution, with lower pH promoting higher iodine evolution. Because the RWST does not experience high doses, water radiolysis is not a direct contributor to the evolution of iodine. The BNP-100 study was based on acidic solutions not including the effects of radiolysis, more applicable to KPS's RWST.

Therefore, the partition coefficient curves from the BNP-100 study are directly applicable to this configuration.

Dominion's application of the iodine partition coefficient model to calculate releases from back-leakage to KPS's RWST during the recirculation phase of the LOCA accident contains assumptions which provide conservatisms that yield higher than expected dose consequences from this release pathway. The total iodine concentration in the RWST water is the critical factor that determines the partition coefficient.

The concentration of iodine in the RWST increases over time with the maximum occurring at the end of the 30 days. The concentration of iodine in the RWST at the end of 30 days is about 3 milligrams/liter.

The partition coefficient (PC) will be at a minimum when the concentration of iodines in the RWST is the greatest at the end of 30 days. The smaller the PC, the larger the amount of iodine that will partition from the water phase and enter into the gas phase and be available for release.

The PC corresponding to the maximum iodine concentration of about 3 milligrams/liter was taken from Figure 8 of the BNP-100 study.

The PC is approximately 600.

Using the equation from SRP 6.5.2, Revision 2 (Reference 15), and the ratio of RWST liquid to air at the end of 30 days of approximately 0.6, the DF is calculated to be approximately 360.

DF = 1 + (Vliquid / Vair) x PC The absolute lowest calculated DF was calculated to occur around 300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> into the accident based on the effect of varying liquid to air volumes and a decreasing partition coefficient as concentration in the liquid increases with time.

Serial No.12-214 Page 7 of 35 Calculated DF in RWST based on Iodine Partition Coefficient 900 800 ~

700 600 400 300 200 100 0

0.1 3

6 12 24 48 96 200 400 720 Hours Dominion selected to model the release of iodine from the RWST liquid throughout the entire accident using a DF value of 100 for conservatism (more than three times lower than the lowest calculated DF value).

The volumetric flow of gas containing iodine partitioned out of the liquid is released out of the RWST at a rate equal to the liquid volumetric leakage rate into the tank. There is no mechanical phenomenon acting upon the tank to promote an accelerated exchange of gas from within the tank. When compared to a dynamic partition coefficient which would start as high as 5000 and gradually decrease to the minimum value of 600, the integrated curies of iodine released under each modeling scenario would indicate the model selected using a DF=100, releases at least 5 times more iodine than actually anticipated. Even with the application Dominion has selected to model RWST releases with a conservative DF=100, the total contribution to EAB dose from the RWST is only 0.1%. If the DF were decreased to a value of 40, the contribution to EAB dose would increase to only 0.2%.

The KPS RWST is not a significant contributor to offsite or control room dose during a LOCA.

This is the same conclusion that was found in the SE to Amendment 166 (Reference 8).

In summary, the RWST is best described as a nearly closed system that does not experience high temperature, radiation, or forced ventilation. Without the effects of high doses, water radiolysis is not a direct contributor to the evolution of iodine due to lower pH, as discussed in recent studies. The primary behavior of iodine in the RWST liquid will be to achieve equilibrium conditions with the air in the tank. Since the liquid entering the tank and the water in the tank are both less than 212 'F, flashing will not occur. Partitioning will occur to achieve equilibrium conditions in the RWST. The BNP-100 study addresses iodine partitioning in acidic solutions without the effects of radiolysis. The RWST static environment is applicable to the conditions discussed in

Serial No.12-214 Page 8 of 35 the BNP-100 study. DEK calculated the lowest expected DF based on the lowest calculated partitioning coefficient over the entire duration of RWST releases. Using a DF value of 100 for the entire modeled release provides at least a factor of five conservatism compared to the expected integrated release that would result if dynamic partitioning were used over the 30-day release period.

Serial No.12-214 Page 9 of 35

4.

NRC Question 4 (ME711O-RAII-AADB-Blum-004-2012-03-02), page 49 states that whole body dose conversion factors are used in the LOCA dose calculation and that the reason for the change is Regulatory Guide (RG) 1.183.

RG 1.183 does not discuss whole body dose conversion factors.

Please clarify what was changed and justify the change.

DEK Response:

Within the industry, the term 'whole body dose' is essentially synonymous with 'effective dose equivalent' (EDE). The label 'whole body' used in Table 3.2-5 under the 'Dose Conversion Factors' parameter heading should ideally be labeled "EDE".

The current analyses of record are Westinghouse proprietary codes and methods with the exception of the FHA, which is analyzed with RADTRAD-NAI. The Westinghouse TITAN5 code uses dose conversion factors for the 'whole body' derived from ICRP-30 (Reference 4) for noble gases and average disintegration energies for the remainder of the nuclides. The dose conversion factors used in the revised RADTRAD-NAI accident analyses are described in section 1.3.2 of the LAR. The specific discussion F pertaining to the EDE dose conversion factors used is copied here:

"The DDE is nominally equivalent to the Effective Dose Equivalent (EDE) from external exposure if the whole body is irradiated uniformly. Since this is a reasonable assumption for submergence exposure situations, EDE is used in lieu of DDE in determining the contribution of extemal dose to the TEDE. EDE dose conversion factors were taken from Table///. 1 of Federal Guidance Report 12 (Reference 16) per Section 4.1.4 of Regulatory Guide 1.183."

Serial No.12-214 Page 10 of 35

5.

NRC Question 5 (ME7110-RAII-AADB-Blum-005-2012-03-02), page 46 states that negative pressure in the shield building is established within 10 minutes. Please verify that "negative pressure" means > 0.25 inch vacuum water gauge with one shield building ventilation system train operating (consistent with Technical Specification SR 3.6.8.2) or justify any proposed change.

DEK Response:

The "negative pressure" assumed in the accident analysis does mean 2! 0.25 inch vacuum water gauge with one shield building ventilation system train operating.

KPS Technical Specification Surveillance Requirement 3.6.8.2 states:

"Verify the shield building can be maintained at a pressure > 0.25 inch vacuum water gauge in the annulus by one Shield Building Ventilation System train with final flow with the limits of Figure 3.6.8-1 within 120 seconds after a start signal."

The accident analysis conservatively assumes that for the first 10 minutes following an accident, shield building ventilation is not credited.

90% of containment leakage is assumed to be released directly to the environment and 10% released through the auxiliary building.

For the period between 10 minutes and 30 days following an accident, 1% of containment leakage is assumed to bypass the shield building and be released directly to the environment. Therefore, for the period between 10 minutes and 30 days following an accident, the shield building ventilation system is credited.

The KPS USAR, Appendix H.3 describes the drawdown of the shield building using the shield building ventilation system.

With one train of the shield building ventilation system operating, the annulus pressure reaches -1.0 inch water column (WC) 8.9 minutes after the Design Basis Accident.

The duration that the annulus is above atmospheric pressure is 3.2 minutes.

Therefore, analysis shows that shield building pressure will be > 0.25 inch vacuum water gauge in less than 10 minutes with one shield building ventilation system train operating.

Serial No.12-214 Page 11 of 35

6.

NRC Question 6 (ME7110-RAII-AADB-Blum-006-2012-03-02), page 69 states:

The LOCA causes a Safety Injection (SI) signal, which also isolates the control room (per current Licensing Basis). The control room is isolated within 10 seconds after the SI signal. Ba'sed on RG 1.183, the onset of the gap release does not start until 30 seconds post-LOCA. Therefore, the control room will be isolated prior to the arrival of the radioactive release.

Technical Specification 3.4.16, "RCS [reactor coolant system] Specific Activity" allows radioactivity to be present in the RCS prior to an accident.

The above justification, that states the control room will be isolated before the release of radioactivity, does not seem to consider that, by design, radioactivity may be present in the RCS prior to the gap release (at the start of the accident). Please provide a complete justification for not considering the impact of the RCS activity prior to the release of gap activity for both the control room and offsite analyses.

DEK Response:

The impact of the RCS activity prior to the release of gap activity was considered for both the control room and offsite dose analyses. The primary containment vent system was modeled as active at the initiation of the accident, prior to the LOCA gap release phase. RCS coolant radionuclide inventory is released through the primary containment vent system until the primary containment vent system is isolated on receipt of a safety injection (SI) signal. A 90 second delay in the isolation of the primary containment vent system conservatively covers the time necessary to receive the SI signal and close the containment vent system valves. The control room was assumed to not be isolated for the first 63 seconds following the accident. Using a conservative containment vent flow rate of 1300 cubic feet per minute (cfm), the resulting contributions to control room, EAB, and LPZ doses were negligible (less than 2 mRem in each case).

Serial No.12-214 Page 12 of 35

7.

NRC Question 7 (ME7110-RAII-AADB-Blum-007-2012-03-02), page 35 states:

Containment purge isolates within 37 seconds following the LOCA and is an insignificant contributor to control room and offsite dose., page 35 states:

KPS is a licensed leak before-break LBB plant (Reference 9). Per RG 1.183, the onset of gap release can be credited with a 10 minute delay for LBB. Containment purge isolation occurs within 37 seconds. Therefore, dose contribution from only TS RCS inventory is insignificant.

Reference 9 provides the citation for a letter to the NRC. Please provide a reference for the NRC safety evaluation which approved LBB methodology.

Also, please provide a justification for applying the LBB methodology to the control room and offsite dose calculations.

DEK Response:

The safety evaluation approving the application of LBB is documented in a letter from K.

E. Perkins (NRC) to D. C. Hintz (WPS) dated February 16, 1988. This document could not be located in ADAMS. Therefore, a copy of the letter is provided as Enclosure 1 to this letter. In the current submittal, credit for LBB methodology was limited to the LOCA analysis, consistent with the guidance in RG 1.183, Section 3.3 (Reference 7). The impact of crediting the LBB methodology in the LOCA analysis was limited to modeling the onset of the gap release phase after the primary containment vent closes at 37 seconds into the event.

An estimate of the impact of removing the credit for LBB methodology on the resulting control room and offsite doses was made using the same model that was used for the LOCA, but assuming an additional unfiltered flow of 1300 cfm through the primary containment vent for the first seven seconds of the gap release phase (which begins at 30 seconds). The increase in the control room and offsite doses was less than 7 mRem. Thus, the impact of crediting the LBB methodology in the LOCA analysis is negligible.

Serial No.12-214 Page 13 of 35

8.

NRC Question 8 (ME711O-RAII-AADB-Blum-008-2012-03-02)

Please explain why switchover to recirculation spray is not credited in the LOCA analysis. Also, please state if any operator action is credited in the assumption that the RWST switchover occurs at 0.91 hours0.00105 days <br />0.0253 hours <br />1.50463e-4 weeks <br />3.46255e-5 months <br />. If operator actions are credited please provide the NRC staff safety evaluation where these operator actions are approved for the design basis LOCA dose calculation.

DEK Response:

The Internal Containment Spray (ICS) system provides clean spray from the RWST to containment atmosphere during a design basis LOCA radiological analysis. The source of water for spraying recirculated sump fluid inside containment is provided by aligning an operating Residual Heat Removal (RHR) pump to the suction of an operating ICS pump. KPS credits manual realignment of RHR pump suction to the sump after the RWST has been depleted. Recirculation spray is not credited in response to a design basis accident at KPS. License Amendment 184 to the KPS Technical Specifications was issued by the Nuclear Regulatory Commission on June 21, 2005 (Reference 12).

The amendment removed the requirement for the ICS pumps to draw suction from the containment sump via the RHR pumps and supply containment spray when in the containment sump recirculation operating mode.

This change was made due to concerns with runout of the RHR and/or ICS pumps when the RHR pumps are supplying suction to the ICS pumps in the recirculation mode. Therefore, no credit is

.taken in the analyses (containment and radiological) for operation of ICS during the recirculation phase of the LOCA.,

No operator action is credited to terminate ICS flow at 0.91 hours0.00105 days <br />0.0253 hours <br />1.50463e-4 weeks <br />3.46255e-5 months <br /> in the LOCA radiological analysis. Ending containment spray early is conservative to dose because it results in an increase in the release from containment to the environment from that time forward by ending one of the removal mechanisms.

0.91 hours0.00105 days <br />0.0253 hours <br />1.50463e-4 weeks <br />3.46255e-5 months <br /> represents the earliest time the RWST could reach the level where operator action would be initiated to end ICS flow based on operation of one train of ECCS.

Serial No.12-214 Page 14 of 35

9.

NRC Question 9 (ME7110-RAII-AADB-Blum-009-2012-03-02), page 29 states:

As a result of the analyses documented in this LAR, the alternate control room intake will be restricted from use. This restriction is required because of the X/Q that would result due to the close proximity of the alternate intake to various release points; one of which is < 10 m from the alternate intake. Administrative controls will be in place to assure the alternate control room intake is closed and prohibit its use during normal operation, following an accident, or while moving recently irradiated fuel. [emphasis added]

Regulatory Position (RP) 5.1.2 of RG 1.183 states:

5.1.2 Credit for Engineered Safeguard Features Credit may be taken for accident mitigation features that are classified as safety related, are required to be operable by technical specifications, are powered by emergency power sources, and are either automatically actuated or, in limited cases, have actuation requirements explicitly addressed in emergency operating procedures. The single active component failure that results in the most limiting radiological consequences should be assumed.

Assumptions regarding the occurrence and timing of a loss of offsite power should be selected with the objective of maximizing the postulated radiological consequences.

The licensee's response to Question 5, of a letter dated November 8, 2011 (Adams Accession No. ML11318A205) provides proposed changes to limiting condition for operation (LCO) 3.7.10 to ensure the Control Room Envelope is isolated during movement of recently irradiated fuel. The response to Question 5 and the license amendment submittal do not appear to provide a similar provision, consistent with RP 5.1.2 cited above, for ensuring that the alternative control room intake is closed during normal operation or following an accident.

RG 1.183, RP 5.1.2 provides credit for mitigation features that are required to be operable by technical specifications. Justify why credit for isolation of the alternative control room intake is assured (during normal operations or following an accident) by.

the proposed methods or propose a method consistent with RP 5.1.2.

DEK Response:

The Control Room Ventilation system is operated under Normal Operating Procedure OP-KW-NOP-ACC-001, "Control Room Air Conditioning System."

The current procedure allows use of either the normal or alternate air intake although the normal air intake is normally aligned. This procedure will be revised, as part of the implementation

Serial No.12-214 Page 15 of 35 process for this amendment, to only allow the normal air intake to be aligned. The alternate intake will be required to be maintained shut.

The KPS control room is a neutral pressure control room during design basis events.

The control room completely isolates.

A Safety Injection, High Radiation or Steam Exclusion signal results in both the normal and alternate intakes to shut, if open.

Procedure AOP-ACC-001, "Abnormal Control Room A/C System Operation," provides guidance for system operation under design basis accident conditions. The normal and alternate air intakes are verified to be shut if a valid SI, High Radiation or Steam Exclusion signal exists. This procedure allows operators to open either intake for a short duration to provide fresh air to the control room.

The fresh air is introduced through the control room charcoal and HEPA filters. This procedure will also be revised as part of the implementation process for this amendment to not allow opening of the alternate intake.

There are also additional procedures identified that allow opening the alternate intake that will require revision as part of the implementation process for this amendment:

• AOP-FP-003, "Fire in a Dedicated Fire Zone"

" OP-KW-NOP-RM-003, "Control Room Radiation Monitor Functional Checks" In summary, isolation of the control room alternate intake during normal operation and following an accident will be assured by procedural controls. Appropriate procedures will be changed during the implementation process for this amendment so that opening of the alternate intake is not allowed.

The design of the alternate intake dampers supports the conclusion that these dampers will remain closed at all times. The alternate intake dampers are redundant safety-grade air-operated isolation dampers. They are designed to close upon receipt of a safety injection, steam exclusion / zone SV signal or a loss of offsite power based on the operation of a spring.

DEK believes that the combination of procedural controls requiring the alternate intake to be maintained in a closed position and the design of the dampers will assure that the dampers will remain closed before and after an accident.

Serial No.12-214 Page 16 of 35

10. NRC Question 10 (ME7110-RAII-AADB-Blum-010-2012-03-02), page 68 states:

Flows reduced from nominal values... by a factor equal to the inverse of the partition coefficient derived from a DF of 100.

Why are the flows reduced by the partition factor rather than by using the decontamination factor (DF)? How are the DF and partition coefficient defined for this application (based upon volume or mass)?

DEK Response:

The use of a partition coefficient (PC) rather than a decontamination factor (DF) facilitated the use of a variable-volume liquid space model for the RWST. The liquid volume of the RWST, and the concentration of iodine in the liquid volume, were modeled as varying with time. The use of a partition coefficient permitted the iodine concentration in the air volume to be directly calculated from the iodine concentration in the liquid volume. The air volumetric flow rate from the RWST (equal to the liquid volumetric inflow rate) then determined the iodine release rate.

As described in the response to RAI 3 (ME7110-RAII-AADB-Blum-003-2012-03-02), the partition factor was calculated from a conservative value of the DF using the equation:

PC = (DF - 1) x (VairNliquid), where; Vair = vapor volume in the tank, and Vliquid = the liquid volume in the tank The conservatism in the calculated partition coefficient was increased by using tank volumes at the end of the event.

Serial No.12-214 Page 17 of 35

11.

NRC Question 11 (ME7110-RAII-AADB-Blum-011-2012-03-02), page B 3.9.6-3 and B3.9.6-4 states:

If it is determined that closure of the equipment hatch and/or containment penetrations would represent a significant radiological hazard to the personnel involved, the decision may be made to forgo [emphasis added] closure of the hatch and/or penetrations.

The above proposed language seems to be contrary to the intent of the TSTF-312

[Technical Specification Task Force],

"Administratively Control Containment Penetrations" [ADAMS Accession No. ML040620147]. TSTF-312 bases approval of the TSTF on whether, the hatch "can and will be promptly closed." Per Title 10 of the Code of Federal Regulations Section 50.67 (10 CFR 50.67), the fission product release assumed for the design calculations are based upon a major accident that results in potential hazards "not exceeded by any accident considered credible."

The need for a provision to forgo closure of the hatch appears to acknowledge that there is a credible scenario where the design source term is exceeded. Therefore, the source term used for the fuel handling accident does not appear to align with 10 CFR 50.67 which specifies the need for a source term that is not exceeded by any credible accident. Please justify the source term used for the fuel handling accident, remove the proposed provision to allow forgoing closure, or provide a source term for the fuel handling accident which is not exceeded by any accident considered credible.

DEK Response:

DEK had proposed adding this additional statement into the TS Bases (discussing an option to not close the equipment hatch under certain radiological conditions) to provide operators with additional flexibility in decision making. However, since there is no credible scenario where the design source term is exceeded, such a statement is not needed. Therefore, the originally proposed statement, regarding the proposed provision to allow forgoing closure, will be removed from the TS 3.9.6 Bases. The originally proposed statement to be deleted from the TS 3.9.6 Bases is provided below.

"If it is determined that closure of the equipment hatch and/or containment penetrations would represent a significant radiological hazard to the personnel involved, the decision may be made to forgo closure of the hatch and/or penetrations."

Serial No.12-214 Page 18 of 35

12. NRC Question 12 (ME7110-RAII-AADB-Blum-012-2012-03-02)

For several design basis accidents the assumed time to isolate the control room is decreased (i.e., Attachment 4, Table 3.2-5, page 58).

For example, the Loss of Coolant Accident assumes the control room isolation damper takes 10 seconds to close upon receipt of a safety injection signal, but the proposed value for control room isolation is zero seconds.

Explain and justify why the proposed value for control room isolation for some accidents is less than the 10 seconds to close the isolation damper.

The revised time to isolate the control room for all accidents does not seem to include the time to start and load the diesel generators. Please justify that, given the worst case single failure, the isolation of the control room does not require diesel power and that the time to isolate the control room is not influenced by time to start and load the diesel.

DEK Response:

The time to start and load diesel generators does not impact the operation of the control room isolation dampers.

The control room supply and exhaust ductwork contains redundant safety grade isolation dampers. Each control room isolation damper closes within 10 seconds of either a safety injection (SI) signal, steam exclusion / zone SV signal or a loss of power based on the operation of a spring. Closure of the control room isolation dampers does not require the EDG. The time to isolate the control room does not require diesel power and the time to isolate the control room is not influenced by time to start and load the diesel.

The LOCA and the FHA are the only analyses with control room isolation times less than the 10 seconds required for isolation dampers to close. As described in LAR-244, the consequences of the FHA are mitigated by requiring that control room isolation dampers be closed prior to moving recently irradiated fuel.

The LOCA is more complex. As described in LAR-244, Attachment 4, Table 3.2-5, in the LOCA event a SI signal is achieved nearly coincident with the start of the event and the control room isolation dampers are closed 10 seconds later.

The gap release begins 30 seconds after the initiation of the event. Hence the control room is isolated prior to the beginning of the gap and in-vessel releases.

In order to simplify the RADTRAD-NAI model, the control room is modeled to be isolated at the start of the gap release for cases that model the containment, ECCS and RWST back-leakage sources.

However, the description of the LOCA also includes, in LAR-244, Attachment 4, Table 3.2-5, brief mention of releases of TS coolant activity through the potentially open containment purge isolation system to account for a 37 second allowed closing time after receipt of a SI signal for a containment purge isolation valve. This case is not

Serial No.12-214 Page 19 of 35 described in LAR-244 as it resulted in negligible dose consequences, but this case did include a 10 second delay for control room isolation due to damper operation and included a 90 second release to bound the 37 second allowance for containment purge isolation valve closure.

Serial No.12-214 Page 20 of 35

13. NRC Question 13 (ME7110-RAII-AADB-Blum-013-2012-03-02)

The Technical Specification "Ventilation Filter and Testing Program," 5.5.9 states that the High-Efficiency Particulate Air and charcoal absorbers are allowed to have a bypass of 1% by design.

How is the allowed bypass of 1% accounted for in the design basis radiological calculations?

If the bypass is not accounted for in the radiological design calculations please consider it in the design calculations or justify why it need not be considered.

DEK Response:

Technical Specification 5.5.9, "Ventilation Filter Testing Program (VFTP)," requires that inplace testing of high efficiency particulate air (HEPA) filters and charcoal adsorbers shows a penetration and system bypass -< 1.0% when tested in accordance with Regulatory Positions C.5.c and C.5.d, respectively, of Regulatory Guide 1.52, Revision 2, and ANSI N510-1975 at specified system flowrates +/- 10%. Since filter testing was first included in Technical Specifications, the acceptance criteria for inplace testing of filter media has been either removal of >99% of DOP test aerosol or halogenated hydrocarbon refrigerant or the current wording that the surveillance shows a penetration and system bypass < 1.0%.

The HEPA filter efficiency assumed in the accident analyses is 99%. Inplace testing of HEPA filters typically show > 99.95% removal with a test acceptance criterion of >99.2%

for margin to the technical specification limit. The assumption in the accident analyses is consistent with the surveillance requirement in Technical Specification 5.5.9.a that penetration and system bypass is < 1.0% without any safety factor included.

The assumption of 99% HEPA filter efficiency is consistent with the existing design basis accident analyses and has been previously reviewed and approved in the radiological analyses submitted for DEK license amendments 166, 172 and 190 (References 8, 13, 14).

The charcoal adsorber filter efficiency assumed in the accident analyses is 95% for the Shield Building Ventilation System (SBVS) and Auxiliary Building Special Ventilation System (ASVS), and 90% for Control Room Post Accident Recirculation System (CRPARS). These filter efficiency assumptions are periodically verified by laboratory testing in accordance with Technical Specification 5.5.9.c and include a safety factor of

2. The proposed RAAs do not account for 1% system penetration and bypass. The 1%

penetration and system bypass requirement of Technical Specification 5.5.9.b is accommodated in the analyses with regard to charcoal adsorbers as allowed by Generic Letter 83-13.

Generic Letter 83-13 states that 1% penetration and system bypass is applicable when a charcoal adsorber efficiency of 95% (or less) is assumed in the NRC Staff's safety evaluation. The assumed charcoal adsorber filter efficiencies are consistent with the existing design basis radiological accident analyses, which were previously reviewed in KPS license amendments 166, 172 and 190.

Serial No.12-214 Page 21 of 35

14. NRC Question 14 (ME7110-RAII-AADB-Blum-014-2012-03-02), page 63, Figure 3.2-2 provides the effective filter efficiencies for filtered flow through the Auxiliary Building Special Ventilation zone filters. The 50 percent plate-out factor in the Auxiliary Building appears to have been used to derive the organic iodine filter efficiency. When appropriate, plate out of iodine is typically only associated with elemental iodine and is conservatively not assumed for organic iodine. Please justify adjusting the organic iodine nominal iodine filter efficiencies by the 50 percent plate-out factor or remove the credit for organic iodine plate out.

DEK Response:

The RADTRAD cases used to model the ECCS and RWST pathways included a 50 percent plate-out factor for both organic and elemental iodine. The radioiodine available for release to the environment was assumed to be 3 percent organic iodine, consistent with the guidance of RG 1.183. Given the small fraction of organic iodine in the total radioiodine available for release, the impact of the 50 percent plate-out factor for organic iodine was expected to be small. An estimate of the impact of the 50 percent plate-out factor in the ECCS and RWST models was made using the same models that were used for the LOCA, but removing the 50 percent plate-out factor for organic iodine.

Additional RADTRAD cases were run to evaluate the impact of the 50 percent plate-out factor for the ECCS model and the RWST model. Removing the 50 percent plate-out factor for organic iodine resulted in LOCA dose increases to the control room, the EAB, and the LPZ of 19 mRem, 0.3 mRem, and 0.1 mRem, respectively. Therefore, the 50 percent plate-out factor for organic iodine has a negligible effect on the LOCA dose results.

Serial No.12-214 Page 22 of 35

15. NRC Question 15 (ME7110-RAII-AADB-Blum-015-2012-03-02), page 68, Figure 3.2-3 states that the unfiltered flow from the containment sump to the Reactor Water Storage Tank (RWST) is decreased by 50 percent after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. While the containment sump is in direct contact with the containment atmosphere and the pressure of the containment atmosphere may decrease over time, the RWST backleakage may also be influenced by pumps that run during the loss of coolant accident. Please justify the 50 percent decrease in RWST backleakage after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

DEK Response:

Backleakage to the RWST at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the LOCA accident is influenced by containment pressure and by the RHR pump and the ECCS configuration at that time in the event response. At 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post-LOCA, the EGGS is operating in the containment sump recirculation mode with only one Residual Heat Removal (RHR) pump running.

EGOS systems and components that function earlier in the LOCA accident response are not needed at this time (e.g. the internal containment spray (ICS) and the safety injection (SI) systems are not functioning, the IGS and SI pumps are stopped and their associated suction source valves to the RWST are closed). The IGS discharge valves are also closed. However, leakage to the RWST is possible at this time in the event through several leakage paths that are discussed below.

The KPS System Integrity Program (SIP) is required to be established, implemented and maintained per KPS TS 5.5.2, "Primary Coolant Sources Outside Containment."

The SIP requires (in part) that EGGS valves that could potentially contribute to backleakage to the RWST after an accident be hydrostatically leak rate tested during each refueling outage. The SIP is designed to determine leakage for accident EGGS alignments that involve operating IGS and SI systems and SI operation in piggy-back mode with RHR.

Individual valves do not have an allowable leak rate but the total combined leakrate from all the valves must be less than 3 gpm. The leak rate for each valve is measured at a pressure that is greater than or equal to the worst case pressure that would be experienced during an accident.

Any repairs necessary to meet the specified overall leakrate must be accomplished prior to the affected (SI and ICS) systems being returned to an operable status following a refueling outage.

Backleakage to the RWST is possible through the following leakage paths:

1. Leakage to the RWST is possible through the SI-300 A/B and SI-301 A/B valves The SI-300 A/B valves are motor operated valves (MOVs) designed to isolate the RHR pump suction from the RWST during containment sump recirculation. The SI-301 A/B valves are check valves (CVs) designed to prevent back flow from the RHR pump suction to the RWST. Each flow path from the RWST to the respective RHR

Serial No.12-214 Page 23 of 35 pump suction contains one SI-300 motor operated valve and one SI-301 check valve in series.

During the transfer to containment sump recirculation, SI-300A and SI-300B are closed, isolating the RHR pump suction piping from the RWST. Leakage via this path would be driven by containment sump head and containment pressure.

Containment pressure is significantly reduced at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the LOCA compared to the containment pressure early in the event. (According to KPS USAR Table 14.3.5-8, LOCA peak containment pressure is 43.1 psig and at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> it is 10.1 psig.

The pressure against SI-301A and SI-301B check valves 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a LOCA will be minimal in comparison to the RHR system pressure during a normal plant shutdown, when RHR cooling is in operation and these valves are held closed at pressures greater than 200 psig. Due to the low induced pressure against the SI-301 A/B check valves during a LOCA accident and the redundant isolation configuration (i.e. the SI-300 and 301 valves are in series), a specific leakage test of SI-301A and SI-301 B is not performed as part of SIP.

2. Leakage to the RWST is possible through the RHR-299 A/B The RHR-299A/B valves isolate RHR discharge flow from the SI pump suction.

Leakage past the RHR-299A/B valves could result in leakage back to the RWST through the SI pump suction piping (through SI-5A/B) or through the SI pump discharge piping (through SI-208 and SI-209). Leakage through the RHR-299A/B valves is measured by the KPS SIP with the associated RHR pump running at a discharge pressure greater than 200 psig.

Recent SIP measurements for RHR-299A/B show leakage results of < 0.01 gpm.

RHR-299 A/B and SI-5 A/B Leakage Path The RHR-299 A/B and SI-5 A/B valves are motor operated valves in series for the leak path to the RWST. The SI-5 valves are located in the SI pump suction piping from the RWST. Leakage through valves SI-5 A/B is measured by the SIP at a test pressure of 200 psig. Recent SIP measurements for SI-5A/B show leakage results of < 0.1 gpm. The low measured leakage results for RHR-299 A/B and SI-5 A/B and the redundant isolation configuration (RHR-299 and SI-5 valves in series for the leak path to the RWST) support the LOCA radiological accident analysis RWST leakage assumptions.

RHR-299 A/B and SI-208 or SI-209 Leakage Path The RHR-299A/B and SI-208 and SI-209 valves are motor operated valves in series for the leak path to the RWST. The SI-208 and SI-209 valves are located in the SI pump discharge mini-flow recirculation line back to the RWST. Leakage through the SI-208 and SI-209 valves is measured by the SIP at a test pressure of 2478 psig.

Recent SIP measurements for SI-208 and SI-209 show leakage results of < 0.1

Serial No.12-214 Page 24 of 35 gpm. The low measured leakage results from RHR-299ANB and SI-208 or SI-209 and the redundant isolation configuration (RHR-299 and SI-208 and SI-209 in series for the leak path to the RWST) support the LOCA radiological accident analysis RWST leakage assumptions.

3. Leakage to the RWST is possible through the RHR-400A/B valves RHR-400A/B valves are motor operated valves located between the RHR pump discharge and the ICS pump suction piping. The RHR-400 A/B valve leakage is not measured by SIP. The leak path through the RHR-400A/B valves to the RWST involves leakage through the ICS suction piping and the ICS discharge piping.

ICS Suction Piping (RHR-400A/B and ICS-3A/B and ICS-2A/B Leakage Path)

Leakage to the RWST through the RHR-400A/B valves would be though the ICS-3A/B and ICS 2A/B valves. ICS-3A and ICS-3B are check valves that are in series with redundant motor operated valves ICS-2A and ICS-2B. The ICS-2A/B valves are closed during containment sump recirculation.

Check valves ICS-3A/B are included in the SIP and leakage through these valves is measured at a test pressure of 200 psig. Recent SIP measurements for the ICS-3A/B valves show leakage of

< 0.1 gpm. The low measured leakage results for the ICS-3 A/B valves and the redundant isolation configuration (RHR-400, ICS-3, and ICS-2 valves in series for the leak path to the RWST) support the LOCA radiological accident analysis RWST leakage assumptions.

ICS Discharge Piping (RHR-400NAB and ICS-5A/B or ICS-6A/B Leakage Paths)

The leak path through RHR-400A/B valve to the RWST through the discharge piping of the ICS system involves leakage through the ICS-5A/B or ICS-6A/B valves. ICS-5A/B and ICS-6A/B would be closed during containment sump recirculation.

However, these valves are not leak checked. There are two potential leak paths back to the RWST if leakage goes past RHR-400A/B and past ICS-5A/B or ICS-6A/B (all closed valves for this scenario), through the ICS full flow test line or through the ICS recirculation line.

Leak Path 1 -Through ICS Full Flow Test Line (ICS-210A/B)

This leakage path would be through RHR-400A/B (closed valve), ICS-5A/B or ICS-6A/B (closed valves) and then through ICS-210A/B (closed valve).

The ICS-210A/B valves are manual valves that are normally closed and would not be repositioned during an accident. Leakage past ICS-210A/B is measured per the SIP at a test pressure of 500 psig. Recent SIP leak measurements for ICS-21OA/B show leakage of < 0.2 gpm.

Leak Path 2 - Through ICS Recirculation Line (ICS-201 or ICS-202)

Serial No.12-214 Page 25 of 35 This leakage path would be through RHR-400A/B (closed valve), ICS-5A/B or ICS-6A/B (closed valves) and then through either ICS-201 and ICS-202 (both closed valves).

ICS-201 and ICS-202 are air operated valves that receive a closed signal during containment isolation. Leakage past ICS-201 and ICS-202 is measured per the SIP at a test pressure of 500 psig.

Recent SIP leak measurements for ICS-201, ICS-202 show leakage of < 0.51 gpm.

For the two leakage paths identified above through the ICS discharge piping, at least two closed MOVs are in series prior to the valve that is leakage tested per the SIP. Leakage back to the RWST would need to be through three closed valves. In each leakage path, at least one of the closed valves is leakage tested per the SIP.

The valves that are leakage tested may not be the boundary valves for this LOCA scenario and for the conditions at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post-LOCA when operating in the containment sump recirculation mode. SIP is designed to determine leakage for accident ECCS alignments that involve operating ICS and SI systems and SI operation in piggy-back mode with RHR. As discussed above, at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> into LOCA the ICS and SI systems are not functioning and many of the valves in these systems are closed. Even though the SIP test conditions are conservative with respect to the expected conditions and configurations at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post-LOCA the valves tested in recent SIP testing are shown to provide satisfactory leakage results to support the LOCA radiological accident analysis assumptions for RWST leakage at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post-LOCA.

Summary The assumption that backleakage to the RWST is reduced by 50% at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after an accident is supported by the following:

1. The KPS System Integrity Program (SIP) measures valve leakage at conditions that are conservative with respect to conditions 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a LOCA. The SIP requires that ECCS valves that could potentially contribute to backleakage to the RWST after an accident be hydrostatically leak rate tested during each refueling outage.

The SIP is designed to measure leakage for post-accident ECCS alignments that involve operating ICS and SI systems and SI operation in piggy-back mode with RHR. The SIP measured leakage that is used to determine total RWST backleakage is conservative, in that the valve with the highest measured leakage in each leakage path is used, without crediting other valves in the leakage path for reducing leakage. The actual expected backleakage to RWST would be determined by valve with the least leakage in the path, not the valve with the most leakage. The SIP acceptance criterion for backleakage is 3.0 gpm.

Recent SIP measurements show total worst-case leakage through the affected valves is < 1.5 gpm at the bounding test conditions.

2. At 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post-LOCA, the ECCS is operating in the containment sump recirculation mode with only one Residual Heat Removal (RHR) pump running.

ECCS systems and components that function earlier in the LOCA accident

Serial No.12-214 Page 26 of 35 response are not needed at this time (e.g. the internal containment spray (ICS) and the safety injection (SI) systems are not functioning, the ICS and SI pumps are stopped and their associated suction source valves to the RWST are closed).

The ICS discharge valves are also closed.

3. The SIP measured leakage is at a conservative system pressure when compared to the pressure at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post-LOCA.

(For example, in the case of the SI pump discharge side valves, their leakage is measured at SI pressure, which is significantly higher than the pressure during containment sump recirculation when the SI pump stopped and only one RHR pump running.)

4. At 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post-LOCA, containment pressure, which influences leakage in certain leakage paths, is significantly reduced.

5. At 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post-LOCA, each leakage path from the ECCS to the RWST is isolated by multiple closed valves in series, and at least one of the closed valves (with the exception of the SI-300 and 301 valves) is leakage tested per the SIP.

(Note that due to the low induced pressure against the SI-301 A/B check valves during a LOCA accident and the redundant isolation configuration (i.e. the SI-300 and 301 valves are in series), a specific leakage test of SI-301A and SI-301B is not performed as part of SIP.)

A Table Summarizing the SIP Test Results for each of the leakage paths discussed above is provided below.

Serial No.12-214 Page 27 of 35 RWST Backleakaqe Measurements Number of 2009 2011 Closed20901 Leakage Path Summary Valves in Measurement Measurement Leak Path (Test Pressure)

(Test Pressure)

Through SI-301A/B and SI-300A/B

[Check Valve (CV) and Motor Operated Valve 2

N/A N/A (MOV) - Not Measured]

SI-5A SI-5A 0.05 gpm 0.09 gpm Through RHR-299A/B and SI-5AIB 2

(200 psig)

(200 psig)

[All MOVs - Each measured]

SI-5B SI-5B 0.02 gpm 0.06 gpm (200 psig)

(200 psig)

SI-209 SI-208 Through RHR-299A/B and SI-208 or SI-209 2

[All MOVs - Each measured]

0.04 gpm 0.05 gpm (2478 psig)

(2478 psig)

ICS-3A ICS-3A Through RHR-400A/B and ICS-3AIB and 0.10 gpm Ogpm Through 3

(200 psig)

(200 psig)

ICS-2AIB

[MOVs and CVs - Only CV's Measured]

ICS-3B ICS-3B 0.10 gpm 0.05 gpm (200 psig)

(200 psig)

ICS-210A ICS-210A Through RHR-400A/B and ICS-5AIB or 0.05 gpm 0.05 gpm ICS-6AIB and ICS-210A/B (500 psig)

(500 psig)

[MOVs and Manual Valves - Only Manual Valves ICS-210B ICS-210B Measured]

0.10 gpm 0.15 gpm (500 psig)

(500 psig)

Through RHR-400A/B and ICS-5AIB or ICS-201 ICS-201 ICS-6AIB and ICS-201 or ICS-202 0.50 gpm 0.25 gpm

[MOVs and AOVs - Only AOVs Measured]

(500 psig)

(500 psig)

TOTAL 0.96 gpm*

0.70gpm*

(<1.5 gpm)

(<1.5 gpm)

• This value represents the sum of the highest measured leakage value for each of the potential backleakage paths to the RWST, without credit for other valves in the leakage path which may have lower leakage. The leak rate for each valve is measured at a pressure that is greater than or equal to the worst-case pressure that would be experienced during an accident.

Serial No.12-214 Page 28 of 35

16. NRC Question 16 (ME7110-RAII-AADB-Blum-016-2012-03-02), page 73 states:

The core curies include a 6% increase to account for fuel management variations (493.6 +/- 10% EFPD [effective full power days], average enrichment of 4.5 w/o +

10%, and core mass of 49.1 MTU [metric ton uranium] +/- 10%).

State whether the burnup of an assembly is limited to 493.6 + 10% EFPD. If not, justify why the assumed burnup is conservative for the fuel handling accident given the fuel is allowed to achieve higher burnups.

DEK Response:

The stated limit of 493.6 +/- 10% EFPD is a core average cycle burnup limit. This limit, along with the stated enrichment and MTU limits, is used in the development of the core radionuclide inventory. All three limits are verified on a cycle specific basis as part of the reload process.

In order to determine the FHA source, the core inventory is adjusted by radial peaking factors from the Core Operating Limits Report (COLR) and gap fractions as directed by RG 1.183, Regulatory Positions 3.1 and 3.2, respectively, except that 25% of the rods in the limiting assembly are modeled as exceeding the RG 1.183, footnote 11 requirement (6.3 kW/ft peak rod average power for burnups exceeding 54 GWD/MTU).

Fuel rods exceeding the footnote 11 requirement of RG 1.183 were modeled at higher gap fractions that were previously approved in KPS License Amendment 190 (Reference 14) (see LAR 244, Attachment 4, Table 3.3-1).

The percentage of rods exceeding the RG 1.183, footnote 11 is verified on a cycle specific basis as part of the reload process.

This is the same method that was used to determine the source term for the FHA as previously approved in Amendment 190 with two exceptions.

In Amendment 190 the core curies included a 3% instead of a 6% increase to account for fuel management variations and the fuel rods exceeding RG 1.183, footnote 11 was 50% instead of 25%.

The change from 3% to 6% increase to account for fuel management variations allows the LOCA, FHA, LRA and REA events to start with the same core inventory and EFPD range. Previously, the LRA and REA analyses were limited to 493.6 +/- 5% EFPD based on the core inventory calculation with the 3% increase.. The change from 50% to 25%

of the rods exceeding footnote 11 simply removes excess conservatism from the analysis based on verification of the percentage of rods exceeding the RG 1.183 footnote 11 on a cycle specific basis as part of the reload process.

Serial No.12-214 Page 29 of 35

17. NRC Question 17 (ME7110-RAII-AADB-Blum-017-2012-03-02), page 84 states:

Based on the assumption that the fuel assembly will be horizontal once it comes to rest, it was determined that an assembly lying on the reactor vessel flange will have approximately 22.35 feet of water above the highest point of the assembly to the water surface. In the spent fuel pool, greater than 23 feet of water will exist.

The depth of 22.35 feet of water was evaluated to verify an effective decontamination factor of 200 using WCAP-7828 (Reference 27).

Using the methods defined in the WCAP with conservative assumptions to minimize predicted decontamination factors for various depths of water, a DF of greater than 500 was determined for elemental iodine. The use of an overall effective DF of 200 was determined to be appropriate per RG 1.183.

Justify the assumption that the fuel assembly will be horizontal once it comes to rest for the accident in the containment. Are there any obstructions which could prevent it from becoming horizontal?

Provide all the input assumptions and the methodology used in the determination that a DF of 200 with less than 23' feet of water is appropriate. Justify why WCAP-7828 is appropriate for the fuel used at your facility.

DEK Response:

To address the questions in this RAI regarding the impact of less than 23 feet of water, Dominion has modeled the generally-accepted Burley method as well as the WCAP-7518 method to demonstrate that the two models result in effective DFs that are greater than that recommended in Regulatory Guide 1.183 (DF = 200). In this assessment, 22 feet of water will be modeled to show the conservatism in using an effective DF of 200.

Based on a drawing review, the tallest obstructions which might prevent a fuel assembly from assuming the horizontal position is the compression plate, used for the cavity seal, which projects approximately two inches above the reactor flange elevation. The top of the compression plate stands approximately. two inches above the flange and circumferentially covers the annulus gap between the reactor vessel flange and the reactor cavity. There are toggle bolts on the plate that stand about 3.5 inches above the reactor vessel flange elevation but due to their size (estimated horizontal dimensions of 2.5" x 2.5") represent a highly unlikely obstruction. The compression plate is a relatively large obstruction which should be credited.

Serial No.12-214 Page 30 of 35 With a horizontal fuel assembly height of 7.8 inches, the peak height of the assembly with the additional two inches from the compression plate is approximately 10 inches above the reactor vessel flange. Since Technical Specifications require a minimum of 23 feet of water above the reactor vessel flange, this leaves only 22.16 feet of water above the fuel assembly.

As a result of this, Dominion will use 22 feet to ensure conservatism in its assessment.

WCAP-7518, "Radiological Consequences. of a Fuel Handling Accident," June 1970, provides supporting information for WCAP-7828 and for the Burley method (Reference B-1 of Regulatory Guide (RG) 1.183) which is used to devel op the approved effective DF of 200 discussed in RG 1.183. Both WCAP-7518 and the Burley Report will be used in the response below, not WCAP-7828.

Per Appendix B of RG 1.183, the following information is provided:

1. Iodine released from the fuel to the pool is assumed to be 95% cesium iodide (CsI), 4.85% elemental iodine and 0.15% organic iodide.

The iodine from Csl dissociates and re-evolves as elemental.

2. For a water depth of 23 feet, the decontamination factors (DFs) for elemental and organic species are 500 and 1, respectively.

3. An overall effective DF of 200 is the result.

It has been determined that there is only 22.16 feet of water above the fuel. In this assessment 22 feet will be conservatively assumed.

After reviewing the Burley method and WCAP-7518, Dominion, concludes that the Burley method assesses bubble diameter based on an analytical method and conservative assumptions. WCAP-7518 assesses bubble diameter based on empirical data from actual small and large scale tests, and as a result provides an effective (average) bubble diameter taking into consideration release orifice size and release pressure. This response primarily addresses bubble diameters used in the calculation of deep water DFs utilizing the Burley method and WCAP-7518. The discussion that follows will show that the WCAP and the Burley method provide results that show use of an effective DF of 200 is conservative.

The WCAP-7518 Approach Dominion is proposing that information provided in WCAP-7518 can be used to determine an effective DF based on KPS fuel rod characteristics.

This WCAP has applicability to the KPS design because test conditions are very similar to KPS with regard to orifice size (fuel rod internal diameter), release pressure (rod internal pressure) and water depth. The resulting effective DF will show that use of a DF of 200 per RG 1.183 is bounding.

Serial No.12-214 Page 31 of 35 The WCAP-7518 method uses the equation:

DF = 73e° 3 0

3 t/d where; t = bubble rise time (sec) d= bubble diameter (cm)

KPS fuel rods have an inner diameter of 0.374 inches. This represents the orifice that bubbles will exit from. Test data for a 23 foot water depth in Table 3-5 of WCAP-7518 show an effective bubble diameter of 0.710 cm for a 1200 psig release pressure from a 0.37 inch orifice, and minimum bubble rise time of 4.7 seconds. Also from Table 3-5, a low pressure (100 psig) release results in an effective bubble diameter of 1.01 cm but a longer rise time. This bubble diameter is 42% larger than the effective diameter from a 1200 psig release. Assuming a maximum effective bubble diameter of 1.01 cm coupled with a minimum rise time of 4.7 seconds results in a conservative low elemental (inorganic) DF of 313 based upon the above equation in 23 feet of water.

The effective DF is determined in the same manner as the Burley method; 1

DFeff =

[(fraction inorganic/DFinorg)+ (fraction organic/I)]

where; fraction inorganic =

0.9985, from elemental and particulate fraction organic =

0.0015, from organic DFinorg represents the elemental DF This results in an effective DF of 213 which is greater than the DF of 200 in RG 1.183.

To address the issue of reduced water height of 22 feet, the primary difference is bubble rise time. From the previous paragraph, bubble rise time for 23 feet is 4.7 seconds.

This is equivalent to an average velocity of 149 cm/ sec. With this velocity, the rise time over 22 feet is 4.5 seconds. Using the same conservative bubble diameter of 1.01 cm results in an elemental DF of 294, and an effective DF of 204. Based on this, the use of an effective DF of 200 is appropriate.

A summary of elemental and effective DFs using WCAP-7518 is provided in Table 1.

Table 1: Parameter Change WCAP-7518 WCAP-7518 (bubble diameter and rise height)

(elemental DF)

(effective DF)

WCAP-7518 Baseline (0.71 cm diameter and 23 feet) 579 310 Burley Baseline (1.6 cm diameter and 23 feet) 183 143 1.01 cm diameter and 23 feet 313 213 1.01 cm diameter and 22 feet 294 204

Serial No.12-214 Page 32 of 35 The Burley Method The Burley method uses an analytical approach to determine a conservative bubble diameter.

It factors multiple conservatisms (large bubble size and a small partition factor) into its determination of DF.

Using the Burley method for 23 feet of water height results in an elemental DF of 133 and an effective DF of 100 as discussed on page 26 of RG 1.183, Reference B-1 (it should be noted that an elemental DF of 133 results in an actual effective DF of 111).

Applying the same method for 22 feet of water results in an elemental DF of 108 with a resultant effective DF of 93.

All factors (e.g., bubble velocity, bubble size, partition factor, etc.) remain the same since the difference of 1 foot will not significantly change the bubble characteristics. These DFs are extremely conservative relative to the DF of 200 provided in RG 1.183.

The Burley method uses the equation; DFinorg = e(-)(keff)(b) where; db = bubble diameter (cm) kerr = effective mass transfer coefficient H = rise height (cm)

Vb

= bubble velocity (cm/ sec)

While this report is not fully specific in the parameters used to determine its conclusion of an inorganic DF of 133, by applying the cited 1.6 cm bubble diameter size and partition factor of 10, and averaging the laminar and turbulent flow DFs results in an inorganic DF of 133.

Applying the more realistic, but conservative bubble diameter of 1.01 cm (discussed above) to the Burley method results in an elemental DF of 26000 and is significantly greater than the WCAP elemental DF of 313. For 23 feet of water the resulting effective DF is 650. On this basis, the use of an effective DF of 200 for 23 feet is conservative.

Adjusting the parameters in the Burley method to reflect a 1.01 cm bubble and 4.5 second rise time (based on 22 feet of water) results in an elemental DF of 15000. The resulting effective DF is 638.

On this basis, the use of an effective DF of 200 is conservative for a 22 foot water height.

A summary of elemental and effective DFs using the Burley method is provided in Table 2 below.

Serial No.12-214 Page 33 of 35 Table 2: Parameter Change Burley Burley (bubble diameter and rise height)

(elemental DF)

(effective DF)

WCAP-7518 Baseline (0.71 cm diameter and 23 feet) 7.2E+07 666 Burley Baseline (1.6 cm diameter and 23 feet) 133 100*

1.01 cm diameter and 23 feet 2.6E+04 650 1.01 cm diameter and. 22 feet 1.6E+04 640

• The calculated value is actually 111 using RG 1.183 iodine composition but reported in Burley as 100 based on assumption of 0.25% organic iodine.

Conclusion The evaluated effective DFs are summarized in Table 3.

Table 3: Parameter Change Burley WCAP-7518 (effective DF)

(effective DF)

WCAP-7518 Baseline (0.71 cm diameter and 23 feet) 666 310 Burley Baseline (1.6 cm diameter and 23 feet)*

100 143**

1.01 cm diameter and 23 feet 650 213 1.01 cm diameter and 22 feet 640 204 The Burley method utilizes conservative assumptions (large bubble diameter and small partition coefficient) to drive the resulting elemental DF to a bounding, conservatively small value of 133.

• • It should be noted that the WCAP method (based on experimental results) using the Burley assumptions of 1.6 cm diameter and 4.7 second rise time result in an elemental DF of 183 and an effective DF of 143.

In conclusion, DEK has evaluated the effect of less than 23 feet of water over a damaged fuel assembly. Parametric examination of techniques used in WCAP-7518 studies and the Burley Report show that the bubble diameter of gases escaping from a damaged assembly effects the measured and analyzed DF. Dominion utilized realistic, yet conservative bubble diameter measurements obtained from actual test conditions that match conditions that would be experienced by a FHA at KPS. Effective DFs were calculated using both the WCAP and Burley method. Use of an effective DF of 200, as provided in RG 1.183, is shown to be conservative for a 22 foot water height at KPS.

Serial No.12-214 Page 34 of 35 References

1.

Letter from J. A. Price (DEK) to Document Control Desk (NRC), "License Amendment Request 244, Proposed Revision to Radiological Accident Analysis and Control Room Envelope Habitability Technical Specifications," dated August 30, 2011. [ADAMS Accession No. ML11252A521]

2.

Letter from Craig W. Lambert (NMC) to Document Control Desk (NRC), "Generic Letter 2003-01; Control Room Habitability - Supplemental Response," dated April 1, 2005. [ADAMS Accession No. ML050970303]

3.

E-mail from Karl D. Feintuch (NRC) to Craig D. Sly, Jack Gadzala (DEK), "ME71 10 Kewaunee Amendment Request Re: Chi-over-Q - AADB Request for Additional Information (RAI)," dated March 2, 2011. [ADAMS Accession No. ML12066A008]

4.

International Commission on Radiological Protection, Publication 30, "Limits for Intakes of Radionuclides by Workers," 1979.

5.

Federal Guidance Report No. 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1988.

6.

TSTF-490, Revision 1, "Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec," dated March 14, 2011.

[ADAMS Accession No.

ML110730473]

7.

NRC Regulatory Guide

)1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," dated July 2000.

8.

Letter From John Lamb (NRC) to Tom Coutu (NMC), "Kewaunee Nuclear Power Plant - Issuance of Amendment Regarding Implementation of Alternate Source Term (TAC No. MB4596)," dated March 17, 2003.

[ADAMS Accession No. ML030210062] (Amendment 166)

9.

Letter from V. Nerses (NRC) to D. A. Christian (DNC), "Millstone Power Station, Unit No. 3 - Issuance of Amendment RE: Alternate Source Term (TAC NO.

MC3333)," dated September 15, 2006. [ADAMS Accession No. ML0619901351

10. Letter from S. P. Lingam (NRC to D. A. Christian (VEPCO), "North Anna Power Station, Unit Nos. 1 and 2, Issuance of Amendments Regarding Technical Specification Changes Per Generic Safety Issue (GSI) 191 (TAC NOS. MD3197 and MD3198)," dated March 13, 2007. [ADAMS Accession No. ML070720043]

11.

Letter from W. R. Matthews (DNC) to NRC Document Control Desk, "Response to Request for Additional Information Regarding Proposed Technical Specification Changes for Implementation of Alternate Source Term," dated March 23, 2005

[ADAMS Accession No. ML050950215]

12. Letter from C. F. Lyon (NRC) to M. G. Gaffney (DEK), "Kewaunee Nuclear Power Plant - Issuance of Emergency Amendment Regarding Containment Spray Flow Path (TAC No. MC7335)," dated June 21, 2005 [ADAMS Accession No. ML051710030]

Serial No.12-214 Page 35 of 35

13. Letter from John Lamb (NRC) to Tom Coutu (NMC), "Kewaunee Nuclear Power Plant - Issuance of Amendment Regarding Stretch Power Uprate (TAC No.

MB9031)," dated February 27, 2004.

[ADAMS Accession No. ML040430633]

(Amendment 172)

14. Letter from R. F. Kuntz (NRC) to D. A. Christian (DEK), "Kewaunee Power Station

- Issuance of Amendment RE: Radiological Accident Analysis and Associated Technical Specifications Change (TAC No. MC9715)," dated March 8, 2007.

[ADAMS Accession No. ML070430017] (Amendment 190)

15. NUREG 0800, "NRC Standard Review Plan," Section 6.5.2, "Containment Spray as a Fission Product Cleanup System," Revision 2, December 1988.

Serial No.12-214 ENCLOSURE 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION:

LICENSE AMENDMENT REQUEST 244, PROPOSED REVISION TO RADIOLOGICAL ACCIDENT ANALYSIS AND CONTROL ROOM ENVELOPE HABITABILITY TECHNICAL SPECIFICATIONS NRC REQUEST FOR ADDITIONAL INFORMATION QUESTIONS AND DOMINION ENERGY KEWAUNEE RESPONSES Contents:

Copy of Letter from K. E. Perkins (NRC) to D. C. Hintz (WPS) dated February 16, 1988 KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE INC.

NUCLEARUNITED STATES

-7io 0 (a

NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 "4,w

,February 16, 1988 Docket No.

50-305 Mr. 0. C. Hintz Vice President - Nuclear Power Wisconsin Public Service Corporation Post Office Box 19002 Green Bay, Wisconsin 54307-9002

Subject:

APPLICATION OF LEAK-BEFORE-BREAK TECHNOLOGY AS A BASIS FOR KEWAUNEE NUCLEAR POWER PLANT STEAM GENERATOR SNUBBER REDUCTION On June 22, 1987, you submitted a request to allow the application of "leak-before-break" (LBB) technology as a basis for a reduction in the number of steam generator (SG) upper lateral support hydraulic snubbers from four to one at each of the two steam generators at the Kewaunee Nuclear Plant.

Your letter of November 20, 1987 transmitted a summary report of the Kewaunee SG snubber reduction analysis for NRC review and approval.

Revised General Design Criterion 4 (GDC-4),

10 CFR Part 50, Appendix A states, "...the dynamic effects associated with postulated pipe ruptures of primary coolant loop piping in pressurized water reactors may be excluded from the design basis when analyses demonstrate the probability of rupturing such piping is extremely low under design basis conditions."

The staff has reviewed WPSC's submittals and finds the Kewaunee primary loop piping is in compliance with revised GDC-4.

Thus, the dynamic effects of postulated primary loop ruptures may be eliminated from the design basis of Kewaunee.

The staff's safety evaluation is enclosed.

Although the Commission has approved the application of LBB technology to eliminate the dynamic effects of postulated primary loop pipe ruptures from the design basis of Kewaunee, a reduction in the number of steam generator (SG) upper lateral support hydraulic snubbers is an unreviewed safety question.

The Commission's response to Issue 11 (52 FR 41291) in the Issues Analysis section of the revised GDC-4 rule states, removal of a pipe whip restraint which also serves as a seismic restraint would not be a 'simple' removal of a pipe whip restraint and, therefore, would involve an unreviewed safety question."

Therefore, it is necessary that WPSC submit an appropriate license amendment application.

Mr. D. C. Hintz Because Kewaunee technical specifications (i.e., technical specification 3.14) do not require snubber operability when the reactor is subcritical, the proposed snubber reduction modification may be implemented during the refueling outage beginning on or about March 5, 1988.

Issuance of the license amendment is required, however, before Kewaunee achieves criticality following the refueling outage.

Sincerely, Kenneth E. Perkins, Director Project Directorate 111-3 Division of Reactor Projects - III, IV, V and Special Projects

Enclosure:

Safety Evaluation cc: See next page

.4 Mr. D. C. Hintz Wisconsin Public Service Corporation Cc:

David Baker, Esquire Foley and Lardner P. 0. Box 2193 Orlando, Florida 32082 Stanley LaCrosse, Chairman Town of Carlton Route 1 Kewaunee, Wisconsin 54216 Mr. Harold Reckelberg, Chairman Kewaunee County Board Kewaunee County Courthouse Kewaunee, Wisconsin 54216 Chairman Public Service Commission of Wisconsin Hill Farms State Office Building Madison, Wisconsin 53702 Attorney General 114 East, State Capitol Madison, Wisconsin 53702 U.S. Nuclear Regulatory Commission Resident Inspectors Office Route #1, Box 999 Kewaunee, Wisconsin 54216 Regional Administrator - Region III U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137 Mr. Robert S. Cullen Chief Engineer Wisconsin Public Service Commission P.O. Box 7854 Madison, Wisconsin 53707 William D. Harvey Vice President and Associate General Counsel Wisconsin Power and Light Company 222 West Washington Ave.

P.O. Box 192 Madison, Wisconsin 53701-0192 Kewaunee Nuclear Power Plant

0 UNITED STATES T) 0 NUCLEAR REGULATORY COMMISSION WASHINGTON, 0. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION WISCONSIN PUBLIC SERVICE CORPORATION WISCONSIN POWER AND LIGHT COMPANY MADISON GAS AND ELECTRIC COMPANY KEWAUNEE NUCLEAR POWER PLANT DOCKET NO. 50-305 ELIMINATION OF POSTULATED PRIMARY LOOP PIPE RUPTURES AS A DESIGN BASIS

1.0 INTRODUCTION

By letter dated June 22, 1987, the Wisconsin Public Service Corporation, (the licensee) requested the application of fracture mechanics "leak-before-break" (LBB) technology to eliminate the dynamic effects of postulated primary loop pipe ruptures from the design basis of the Kewaunee Nuclear Power Plant, as permitted by the revised General Design Criterion 4 (GDC-4) of Appendix A to 10 CFR Part 50.

The licensee's request would permit the redesign of the primary loop supports because the dynamic effects of postulated primary loop pipe ruptures would be eliminated from the design basis using LBB.

The licensee indicated that this request would affect an upcoming steam generator large bore snubber reduction program.

The licensee submitted the technical basis for the elimination of primary loop pipe ruptures for Kewaunee in Westinghouse report WCAP-11411, Revision 1 (Reference 1).

By letter dated October 26, 1987, the licensee submitted WCAP-11619 (Reference 2) In response to the staff's request for additional information.

The licensee also referenced Westinghouse reports WCAP-10456 (Reference 3) and WCAP-10931, Revision 1 (Reference 4), which have been reviewed previously by the staff as discussed in References 5 and 6, respectively.

The revised GDC-4 is based on the development of advanced fracture mechanics technology using the LBB concept.

On October 27, 1987, a final rule was published (52 FR 41288), effective November 27, 1987, amending GDC-4 of Appendix A to 10 CFR Part 50.

The revised GDC-4 allows the use of analyses to eliminate from the design basis the dynamic effects of postulated pipe ruptures in high energy piping in nuclear power units.

The new technology reflects an engineering advance which allows simultaneously an increase in safety, reduced worker radiation exposures, and lower construction and maintenance costs.

Implementation permits the removal of pipe whip restraints and jet impingement barriers as well as other related changes in operating plants, plants under construction, and future plant designs.

Containment design and emergency core cooling requirements are not influenced by this modification.

h-I-Usina the criteria in Reference 7, the staff has reviewed and evaluated the licereee's subm4ttals for ccnoli'-a with the revised GOC-4.

This Safety EvalI.ation provides the staff's findings.

KEWAUNEE PRIMARY LOOP PIPING The Kewaunee primary loop piping consists of 35-inch, 37-inch, and 33-inch nominal diameter hot leg, cross-over lea, and cold lea, respectively.

The piping material in the primary loops is cast stainless steel (SA-351 CF8M).

The piping is centrifugally cast and the fittings are statically cast.

The welding processes used were submerged arc (SAW),

shielded metal arc (SMAW),

and gas tungsten arc (GTAW).

STAFF EVALUATION CRITERIA The staff's criteria for evaluation of compliance with the revised GDC-4 are discussed in Chapter 5.0 of Reference 7 and are as follows:

(1) The loading conditions should include the static forces and moments (pressure, deadweight, and thermal expansion) due to normal operation, and the forces and moments associated with the safe shutdown earthquake (SSE)., These forces and moments should be located where the highest stresses, coincident with the poorest material properties, are induced for base materials, weldments, and safe ends.

(2) For the piping run/systems under evaluation, all pertinent information which demonstrates that degradation or failure of the piping resulting from stress corrosion cracking, fatigue, or water hammer are not likely, should be provided.

Relevant operating history should be cited, which includes system operational procedures; system or component modification; water chemistry parameters, limits, and controls; and resistance of material to various forms of stress corrosion and performance under cyclic loadinas.

(3) The materials data provided should include types of materials and materials specifications used for base metal, weldments, and safe ends; the materials properties including the fracture mechanics parameter "J-integral" (J) resistance (J-R) curve used in the analyses; and long-term effects such as thermal aging and other limitations to valid data (e.g., J maximum, and maximum crack growth).

(4) A through-wall flaw should be postulated at the highest stressed locations determined from criterion (1) above.

The size of the flaw should be large enough so that the leakage is assured of detection with at least a factor of 10 using the minimum installed leak detection capability when the pipe is subjected to normal operational loads.

(5) It should be demonstrated that the postulated leakage flaw is stable under normal plus SSE loads for long periods of time; that is, crack growth,.if any, is minimal during an earthquake.

The margin, in terms of applied loads, should be at least 1.4 and should be determined by a flaw stability analysis, i.e., that the leakage-size flaw will not experience unstable crack growth even if larger loads (larger than design loads) are applied.

However, the final rule permits a reduction of the marain of 1.4 to 1.0 if the individual normal and seismic (pressure, deadweicht, thermal expansion, SSE, and seismic anchor motion) loads are summed absolutely.

This analysis should demonstrate that crack growth is stable and the final flaw size is limited, such that a double-ended pipe break will not occur.

(6) The flaw size should be determined by comparing the leakage-size flaw to the critical-size flaw.

Under normal plus SSE loads, it should be demonstrated that there Is a margin of at least 2 between the leakage-size flaw and the critical-size flaw to account for the uncertainties inherent in the analyses and leakage detection capability.

A limit-load analysis may suffice for this purpose; however, an elastic-plastic fracture mechanics (tearing instability) analysis is preferable.

STAFF EVALUATION The staff has evaluated the information presented in References I and 2 for compliance with the revised GDC-4.

Furthermore, the staff performed independent flaw stability computations using an elastic-plastic fracture mechanics procedure developed by the staff (Reference 8).

On the basis of its review, the staff finds the Kewaunee primary loop piping in compliance with the revised GDC-4.

The following paragraphs in this section present the staff's evaluation.

(1) Normal operating loads, including pressure, deadweight, and thermal expansion, were used to determine leak rate and leakage-size flaws.

The flaw stability.analyses performed to assess margins against pipe rupture at postulated faulted load conditions were based on normal plus SSE loads.

In the stebility analysis, the individual normal load components were summed algebraically and the seismic loads were then added absolutely.

In the leak rate analysis, the individual normal load components were summed algebraically.

Leak-before-break evaluations were performed for the limiting location in the piping.

(2) For Westinghouse facilities, there is no history of cracking failure in reactor coolant system (RCS) primary loop piping.

The RCS primary loop has an operating history which demonstrates its inherent stability.

This includes a low susceptibility to cracking failure from the effects of corrosion (e.g., intergranular stress corrosion cracking), water hammer, or fatigue (low and high cycle).

This operating history totals over 450 reactor-years, including 5 plants each having over 16 years of operation and 15 other plants each with over 11 years of operation.

(3) The material tensile and fracture toughness properties were provided in References I and 2.

Because there are cast stainless steel piping (and fitting) and associated welds in the Kewaunee vrimary loop, the thermal aging toughness properties of cast stainless steel materials were estimated according to procedures in References 3 and 4.

The material tensile properties were estimated using generic procedures.

For flaw stability evaluations, the lower-bound stress-strain properties were used.

For leakage rate evaluations, the averaqe stress-strain properties were used.

(4) Kewaunee has a RCS pressure boundary leak detection system which is consistent with the guidelines of Regulatory Guide 1.45 such that a leakage of one gallon per minute (qpm) can be detected.

The calculated leak rate through the postulated flaw is large relative to the staff's required sensitivity of the plant's leak detection systems; the margin Is at least a factor of 10 on leakage and is consistent with the guidelines of Reference 7.

(5) In the flaw stability analyses, the margin in terms of load for the leakage-size flaw under normal plus SSE loads exceeds 1.4 and is consistent with the guidelines of Reference 7.

(6) The margin between the leakage-size flaw and the critical-size flaw was also evaluated in the flaw stability analyses.

The margin In terms of flaw size exceeds 2 and is consistent with the guidelines of Reference 7.

STAFF CONCLUSIONS The staff has reviewed the information submitted by the licensee and has performed independent flaw stability computations.

On the basis of its review, the staff concludes that the Kewaunee primary loop piping complies with the revised GDC-4 accordina to the criteria in NUREG-1061, Volume 3 (Reference 7).

Thus, the probability or likelihood of large pipe breaks occurring in the primary coolant system loops of Kewaunee is sufficiently low such that dynamic effects associated with postulated pipe breaks need not be a design basis.

Principal Contributor:

S. Lee, Er!TB Date:

February 16, 1988 REFERENCES (1) Westinghouse Report WCAP-11411, Revision 1, "Technical Bases for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Kewaunee," April 1987, Westinghouse Proprietary Class 2.

(2) Westinghouse Report WCAP-11619, "Additional Technical Bases for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Kewaunee," October 1987, Westinghouse Proprietary Class 2.

(3) Westinghouse Report WCAP-10456, "The Effects of Thermal Aging on the Structural Integrity of Cast Stainless Steel Piping for Westinghouse Nuclear Steam Supply Systems," November 1983, Westinghouse Proprietary Class 2.

(4) Westinghouse Report WCAP-10931, Revision 1, "Toughness Criteria for Thermally Aged Cast Stainless Steel," July 1986, Westinghouse Proprietary Class 2.

(5) Letter from B. J. Youngblood of NRC to M. D. Spence of Texas Utilities Generating Company dated August 28, 1984.

(6) Letter from D. C. Dilanni of NRC to 0. M. Musolf of Northern States Power Company dated December 22, 1986.

(7) NUREG-1061, Volume 3, "Report of the U. S. Nuclear Regulatory Comuission Piping Review Committee, Evaluation of Potential for Pipe Breaks,"

November 1984.

(8)

NUREG/CR-4572, "NRC Leak-Before-Break (LBB.NRC) Analysis Method for Circumferentially Through-Wall Cracked Pipes Under Axial Plus Bending Loadt.," May 1986.