PLA-6194, Proposed License Amendments 285 and 253, Extended Power Uprate Application Health Physics Technical Review, Request for Additional Information Responses

Proposed License Amendments 285 and 253, Extended Power Uprate Application Health Physics Technical Review, Request for Additional Information Responses

ML071420047
Person / Time
Site:	Susquehanna
Issue date:	05/09/2007
From:	Mckinney B Susquehanna
To:	Document Control Desk, NRC/NRR/ADRO
References
PLA-6194, TAC MD3309, TAC MD3310
Download: ML071420047 (28)

Text

Brltt T. McKlnney PPL Susquehanna, LLC Sr. Vice President & Chief Nuclear Officer 769 Salem Boulevard Berwick, PA 18603 Tel. 570.542.3149 Fax 570.542.1504 ' ibi btmckinney@pplweb.com PAY 09. ZOO?

• TM U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop OP 1-17 Washington, DC 20555 SUSQUEHANNA STEAM ELECTRIC STATION PROPOSED LICENSE AMENDMENT NO. 285 FOR UNIT 1 OPERATING LICENSE NO. NPF-14 AND PROPOSED LICENSE AMENDMENT NO. 253 FOR UNIT 2 OPERATING LICENSE NO. NPF-22 EXTENDED POWER UPRATE APPLICATION RE: HEALTH PHYSICS TECHNICAL REVIEW REQUEST FOR ADDITIONAL INFORMATION Docket Nos. 50-387 RESPONSES and 50-388 PLA-6194

References:

1) PPL Letter PLA-6076, B. T. McKinney (PPL)to USNRC, "ProposedLicense Amendment Numbers 285for Unit 1 Operating License No. NPF-14 and 253 for Unit 2 OperatingLicense No. NPF-22 ConstantPressurePower Uprate,"dated October 11, 2006.

2) Letter, R. V. Guzman (NRC) to B. T. McKinney (PPL),

"Requestfor AdditionalInformation (RAI) -

Susquehanna Steam Electric Station, Units 1 and 2 (SSES 1 and 2) -

Extended Power UprateApplication Re:Health Physics TechnicalReview (TAC Nos. MD3309 andMD3310), "dated April 12, 2007.

Pursuant to 10 CFR 50.90, PPL Susquehanna LLC (PPL) requested in Reference 1 approval of amendments to the Susquehanna Steam Electric Station (SSES) Unit 1 and Unit 2 Operating Licenses (OLs) and Technical Specifications (TSs) to increase the maximum power level authorized from 3489 Megawatts Thermal (MWt) to 3952 MWt, an approximate 13% increase in thermal power. The proposed Constant Pressure Power Uprate (CPPU) represents an increase of approximately 20% above the Original Licensed Thermal Power (OLTP).

The purpose of this letter is to provide responses to the Request for Additional Information transmitted to PPL in Reference 2.

The Enclosure contains the PPL responses.

AcD(

Document Control Desk PLA-6194 There are no new regulatory commitments associated with this submittal.

PPL has reviewed the "No Significant Hazards Consideration" and the "Environmental Consideration" submitted with Reference 1 relative to the Enclosure. We have determined that there are no changes required to either of these documents.

If you have any questions or require additional information, please contact Mr. Michael H. Crowthers at (610) 774-7766.

I declare under perjury that the foregoing is true and correct.

Executed on:,;&ý4 Z.., 9007

Enclosure:

Request for Additional Information Responses Copy: NRC Region I Mr. A. J. Blamey, NRC Sr. Resident Inspector Mr. R. V. Guzman, NRC Sr. Project Manager Mr. R. R. Janati, DEP/BRP

Enclosure to PLA-6194 PPL EPU Request for Additional Information Responses

Enclosure to PLA-6194 Page 1 of 16 NRC Ouestion 1:

Section 8.4.2 (page 8-5) of your submittal states that the magnitude of any increase in the fission products in the steam and in the reactor water (resulting from the proposed CPPU) is expected to be small and would be bounded by the current licensed thermal power (CLTP) design basis values. Provide the expected percentage increase in the fission products in the steam and in the reactor water for the proposed CPPU at SSES 1 and 2.

PPL Response:

Fission product activity in the reactor steam consists of noble gases released from the core plus carryover activity from the reactor water. The noble gases released during plant operation result from the escape of minute fractions of the fission products in the fuel rods. An increase in the fractional release is not expected with CPPU though the absolute rate of noble gases in the steam will increase roughly in proportion to the change in power level reflecting the increase in noble gas inventory in the fuel rods themselves.

Fission product activity in the reactor water, like the activity in the steam, results from minute releases from the fuel rods. There is no expectation that isotopic releases from the fuel will increase with CPPU although the absolute values of released isotopes will increase due to the increased fuel rod activity inventory.

SSES has had good fuel performance experience resulting in low levels of fission product activity in the steam and coolant. Using ANSI/ANS 18.1-1999 "Radioactive Source Term for Normal Operation of Light Water Reactors" as a basis, the following percentage increases in fission products in the steam and in the reactor water are estimated based upon the change in thermal power level from 3489 MWt to 3952 MWt and steam flow rate from 14.43 Mlbs/hr to 16.53 Mlbs/hr. These ANSI Standard fission product activity levels are higher than SSES operating experience and remain bounded by the SSES Design Basis.

SSES 3489 MVt 3952 MWt ANSI/ANS Parameter Design (ANSI/ANS (ANSI/ANS 18.1-1999 Basis 18.1-1999) 18.1-1999)  % Change Noble Gases Offgas Release Rate LOE+05 3.04E+04 3.48E+04 14.5 (pjCi/sec) @30 minutes Delay Fission Product radiation sources (pCi/gm) during normal operation in Reactor Water 2.25E+00 1.36E-01 1.47E-01 8.1 Reactor Steam 5.40E-02 7.51 E-03 7.99E-03 6.4

Enclosure to PLA-6194 Page 2 of 16 NRC Ouestion 2:

Section 8.4.2 (page 8-5) of your submittal states that the magnitude of any increase in the activated corrosion product (ACP) production in the coolant (resulting from the proposed CPPU) is expected to be negligible. Section 8.5 (page 8-6) of your submittal states that non-coolant activation (corrosion) products are expected to increase in proportion to the thermal power increase. Describe why the increase in ACPs in the coolant is expected to be negligible while the non-coolant activation (corrosion) products are expected to increase in proportion to the thermal power increase.

PPL Response:

The statement on page 8-5 is misleading. It would have been more appropriate to use the term "impact" instead of the term "magnitude." The production of activated corrosion products is expected to increase in proportion to the thermal power increase.

Non-coolant activation products are formed by the activation of impurities in the coolant and or by corrosion of irradiated system materials. The higher flow rate at the higher power level is expected to result in the formation and activation of additional impurities.

The change in corrosion product concentration based on ANSI/ANS 18.1-1999 methods is estimated at 12.2 percent, which is roughly proportional to the power level increase.

The magnitude of this predicted increase is small in comparison to the design basis values.

SSES 3489 MWt 3952 MWt ANSIJANS Parameter Design (ANSI/ANS (ANSIIANS 18.1-1999 Basis 18.1-1999) 18.1-1999)  % Change Non Coolant Activation Product radiation sources (QCi/gm)in Reactor Water 6.20E-02 2.87E-02 3.22E-02 12.2 Reactor Steam 6.20E-05 2.87E-05 3.22E-05 12.2 Planned improvements in the RWCU and condensate systems (such as upgrading RWCU filter/demineralizer, the addition of a new condensate demineralizer and filter), are expected to enhance cleanup capability and compensate for increases in activated wear and corrosion product production resulting from CPPU. Accordingly, the magnitude of the change in the non-coolant activation product concentration is expected to be less than the 12.2 percent described above.

Enclosure to PLA-6194 Page 3 of 16 NRC Ouestion 3:

Section 8.5 (page 8-6) of your submittal states that the Nitrogen- 16 (N- 16) dose rates from main steam lines and related equipment may increase up to 20% due to the combined effects of increased activation rates and reduced transit decay times. Verify that the expected increase in dose rates from N- 16 does not create new radiation, or high radiation areas around condensate bearing systems/components in the turbine building.

PPL Response:

As indicated in PUSAR Section 8.5, an assessment of the impact of CPPU on radiation shielding (and plant radiation zoning) was performed based upon the CPPU related changes in radiation sources. As was noted, the results of this assessment did not require changes to the existing plant radiation zoning documented in SSES design basis radiation shielding calculations. No new or high radiation areas are anticipated by the increase in N-16 sources around condensate bearing systems and equipment in the turbine building.

Radiation measurements taken during the startup testing for CPPU implementation will verify these results.

NRC Ouestion 4:

Section 8.5 (page 8-7) of your submittal states that the CPPU may result in localized hot spots in areas outside feedwater heater rooms and near drywell penetrations. Provide more detailed information on the expected pre- and post CPPU dose rates in these areas and describe what controls/changes will be implemented to maintain worker doses as low as is reasonably achievable (ALARA) and within the occupational dose limits of Title 10 of the Code of Federal Regulations (10CFR) Part 20.

PPL Response:

Radiation control areas in the turbine building (Elev 699 ft) outside the feedwater heater rooms are currently specified as Zone III (< 15 mr/hr). Measured general area dose rates

(@ 100% power, w/H 2 flow rate @ 90scfm) outside these rooms is typically in the range of 2 to 4 mr/hr. However, contact dose rates (from HP survey data) at the entrance (doors) to these rooms increase to about 100 mr/hr. These doors currently are shadow shielded (w/lead blanket shielding racks).

It is expected that following CPPU implementation, these dose rates may increase locally up to 20%. The current plan is to install additional shielding to offset the increased dose rates due to CPPU [for example potential use of 2 layers of 1ft x 4 ft shielding blankets (15 lbs/ft) on each of the shield racks (32 blankets/rack) in the Unit 1 & 2 A, B, & C feed water heater cells].

Enclosure to PLA-6194 Page 4 of 16 Review of the design basis penetration shielding calculations demonstrated conservatism in the selection of the normal operating design basis source terms used and in the specification of a design target dose rate of 0.5 mr/hr (Zone I) in the general area external to the drywell penetrations. Areas outside penetrations are currently designated either Zone II (< 2.5 mr/hr), III (< 15 mr/hr), or V (>100 mr/hr). No change in the zone dose rates are projected to occur following CPPU implementation.

Radiation measurements, which would identify local hot spots, will be taken during startup testing for CPPU conditions and/or during ongoing operations at those conditions in accordance with normal station practices to determine if there is a need to supplement the radiation control practices near the feedwater heater rooms or drywell penetrations.

NRC Ouestion 5:

Section 8.5 (page 8-7) of your submittal states that radiation surveys of selected areas will be conducted as part of the CPPU startup and test plan to identify areas that may require changes in radiation shielding or zone designation. Provide a listing of these selected areas where you will conduct radiation surveys following the proposed CPPU implementation and describe your criteria for selecting these areas.

PPL Response:

Areas were selected based upon current operating experience and include those areas expected to be impacted by the change in power and increased N- 16 activation levels.

Specifically those radiation areas (RA) near steam-affected rooms will be included in the survey plan to document current dose rates and to verify that rooms, areas, and hallways are posted correctly. Contact and 30 cm dose rates off the outside surface of accessible doors in steam affected areas will be obtained. General area dose rates in hallways and walkways near steam-affected rooms will be surveyed to ensure that posting boundaries are satisfactory. Locations to be surveyed include:

TURBINE BUILDING TB Elevation 762 ft:

Verify dose rates and RA posting boundaries around crane hatch and along west wall (due to "shine" from the Turbine Deck).

TB Elevation 729fi:

Verify dose rates outside Moisture Separator and Steam Seal Evaporator room (SSE) doors. Verify general area dose rates and RA posting boundaries on the Turbine Deck.

Surveys will include dose rates east of the SSE room and the area between the MG Sets (due to "shine" from the TB 716 Main Steam Tunnel).

Enclosure to PLA-6194 Page 5 of 16 TB Elevation 699ftl:

Verify dose rates outside Condenser Bay and Feedwater Heater Cell doors. Verify general area dose rates and RA posting boundaries in hallway and feedwater heater bay alcove.

TB Elevation 6 76ft:

Verify dose rates outside the accessible Condenser Bay door, Reactor Feedpump Turbine, and east SJAE Room doors. Verify dose rates on platforms above Reactor Feedpumps.

Verify general area dose rates and RA boundaries in hallway and around fan plenum in 676 railroad bay. Note: It may be necessary to obtain dose rates inside the Reactor Feedpump Turbine rooms to determine if the doors should be posted as a locked high radiation area (LHRA) or HRA and to determine if ladder-blocks and LHRA postings are needed on the platform ladders.

TB Elevation 656ft:

Verify dose rates outside Condenser Bay, accessible Offgas Pipe Tunnel, and recombiner room doors. Also, verify dose rates inside the recombiner panels in TB 656 hallway.

Verify general area dose rates and RA boundaries in hallways and rooms near steam affected areas on this elevation.

REACTOR BUILDING RB Elevation 749ft:

Verify dose rates outside OB MSIV (749 Wingslab) door. Verify general area dose rates and RA boundary near the door.

RB Elevation 719ft:

Verify dose rates outside MS Pipe Tunnel (719 Wingslab) door. Verify general area dose rates and RA boundary near the door.

RAD WASTE BUILDING R W Elevation 676fti:

Verify dose rates outside Water Removal Skid room doors. Verify general area dose rates and RA boundaries in the hallway near these rooms.

NRC Ouestion 6:

Section 8.5 (page 8-7) of your submittal lists several strategies to control shutdown dose rates. On the basis of experience gained from other boiling water reactors (BWRs) which have implemented power uprates of similar magnitude to the one planned at Susquehanna, and in light of the various strategies that you use to control shutdown dose rates at SSES 1 and 2, describe what impact you expect the proposed power uprate will

Enclosure to PLA-6194 Page 6 of 16 have on the annual collective doses at SSES 1 and 2. Provide an estimate of the occupational dose that will result from the plant modifications that will be needed to support the implementation of the proposed power uprate.

PPL Response:

The expected impact of the proposed power uprate on the annual collective doses at Susquehanna was provided in response to an NRC Request for Additional Information in letter PLA-6172 "Proposed License Amendment No. 285 for Unit 1 Operating License No. NPF- 14 and proposed License Amendment No. 253 for Unit 2 Operating License No. NPF-22 Constant Pressure Power Uprate Application - Supplement" dated March 22, 2007.

Experience gained from other BWRs is of limited value because the dose received is largely dependent on the specific modifications and plant characteristics.

Prior to installation of any modification, PPL performs an ALARA Design Review per SSES procedure "Implementation of ALARA Activities for Modification and Design" to determine the expected installation dose and actions, such as shielding, use of mock-ups etc., that can be taken to mitigate dose received.

CPPU Modifications are planned for implementation in years 2007 through 2010 during outage and non-outage periods. Of the modifications identified to date, approximately 25% have been implemented as of the 2007 Unit 2 refueling outage. The remaining modification packages are in various stages of scoping, planning and/or execution.

Therefore, a detailed breakdown of the expected doses by modification is not readily available. As indicated in PLA-6172, the dose estimate for each year during the CPPU implementation period is approximately 230 person rem based on an estimated 130 for scheduled outage related work and 100 person rem for planned non outage work. During the Unit 2 refueling outage, 154 person rem was recorded to accomplish planned refueling and CPPU modification related tasks. The occupational doses received for a sampling of the CPPU tasks is provided in the Table below.

CPPU Modification Estimated / Actual Dose (person rem)

Feedwater Heater Replacement 0.69 / 0.55 Condensate Pump Replacement 0.30 / 0.74 Turbine Generator Rewind 2.14 / 2.00 Vibration/Acoustic Monitoring 5.61 / 6.60 One of the more significant, recently identified, dose intensive modifications associated with the CPPU will be the replacement of the steam dryers. The steam dryers being replaced will be cut down the centerline. These halves will be placed in steel boxes for shielding purposes and to allow handling and transport of the pieces. The boxes will be

Enclosure to PLA-6194 Page 7 of 16 shielded until they are removed from the refueling floor. The majority of the occupational dose from this project will be received when the old steam dryers are removed from the reactor building, transported, and placed in a shielded storage facility until plant decommissioning. Removal from the building is scheduled to be done after each refueling outage in which a steam dryer is replaced. The time and date of transport will be scheduled to occur when the fewest number of people are on site. The removal of the steam dryers from the refuel floor and placement in the storage facility will incur a "one-time" dose impact in 2008 and in 2009. The radiation dose from these moves will be estimated as detailed project planning and associated dose calculations are completed.

The detailed planning will consider all reasonable efforts to minimize doses to plant personnel and members of the public.

While calculated doses from the transport and storage of the replaced steam dryers have not yet been finalized, the intent is to ensure that those doses will be less than both 10%

of the summed doses shown in the response to Question 11 and less than 10% of the difference between the summed doses shown and the limits of 40 CFR 190.

NRC Ouestion 7:

Explain the rationale behind applying a scaling factor of 1.5 to the vital area CLTP doses to account for CPPU changes in power level (Section 8.5, page 8-8 and Table 8-1).

PPL Response:

The dose scaling factor of 1.5 was selected to conservatively account for the CPPU power level impact on the applicable isotopic source terms, energy groupings, and dose rates as a function of post accident time periods. Salient features of this review and the selection of this scaling factor to support the bounding assessment reflected in the PUSAR submittal are discussed below.

The CLTP doses were developed based on TID 14844 source term releases for a DBA LOCA associated with a core thermal power of 3616 MWt. With CPPU, the DBA LOCA core thermal power is increased to 4032 MWt. The core source terms for CLTP are based on the GE generic three year full power burn equilibrium fission product inventories provided in NEDC-32161 P. The CPPU core sources are based upon the inventories provided in the Susquehanna PUSAR Appendix A.

The applicable sources used in the vital area dose calculations are those associated with post accident contained reactor coolant, suppression pool, and reactor steam sources. Post accident time periods considered: 0, 1, 30, and 60 days.

Representative source geometries considered in the mission dose calculations are shielded and unshielded 10 to 20 inch diameter pipes x 50 ft lengths.

Enclosure to PLA-6194 Page 8 of 16 A comparison of the CLTP to CPPU dose rate @ - 1 ft indicates that the unshielded CPPU dose rates exceed the CLTP dose rates by no more than a factor of 1.2 (120%).

Time CLTP CPPU Source Post-Accident DoseRate DoseRate Ratio Type (Days) (mR/hr) (mR/hr) CPPU/CLTP 0 9.824E+08 .I 19E+09 113.90%

Reactor 1 6.542E+07 7.35 1E+07 112.37%

Coolant 30 4.447E+06 5.170E+06 116.26%

60 2.105E+06 2.522E+06 119.81%

0 5.159E+07 5.783E+07 112.10%

Suppression I 5.688E+06 6.334E+06 111.36%

Pool 30 4.260E+05 4.967E+05 116.60%

60 2.083E+05 2.496E+05 119.83%

0 6.341E+08 7.248E+08 114.30%

Reactor_1 3.561E+07 3.967E+07 111.40%

Reactr30 8.268E+05 9.192E+05 111.18%

Steam 60 4.850E+04 5.605E+04 !115.57%

For the vital area mission dose calculations, the contained post accident radiation sources will be located at distances greater than 1 ft and will typically be housed behind concrete shield walls. With greater distance and shielding, source types containing higher energy sources will have a more significant effect on the post accident dose rates and the integrated doses. In order to quantify this effect, the source type activities as a function of energy are examined for the periods 0, 1, 30, and 60 days.

The following Table shows for each source type and post accident time period, those energy groups where the percentage change exceeded the CPPU/CLTP factor of 1.2.

Reactor Coolant Su ppression Pool Reactor Steam Ratio Energy Energy Ratio Energy Ratio (Mev) CPPUICLTP (Mev) CPPU/CLTP (Mev) CPPU/CLTP Time = 0 days post-accident 0.2 156.19% - 0.2 159.71%

Time = 1 day post-accident 0.2 136.26% - 0.2 153.88%

Time =30 days post-accident 0.015 120.21% 0.015 120.12% 0.15 206.1 1%

0.05 239.85% 0.05 239.70% 0.5 122.31%

0.06 137.29% 0.06 137.12% -

0.2 135.26% 0.2 139.59%

1.0 124.14% 1.0 124.18% -

Time = 60 days post-accident 0.015 120.15% 0.015 120.10% 0.03 130.53%

0.03 130.37% 0.03 126.46% 0.15 193.65%

0.05 315.07% 0.05 314.36% 0.5 1 133.42%

Enclosure to PLA-6194 Page 9 of 16 Reactor Coolant Suppression Pool Reactor Steam Ratio Energy Energy Ratio Energy Ratio (Mev) CPPU/CLTP (Mev) CPPU/CLTP (Mev) CPPU/CLTP 0.06 122.08% 0.06 121.99% -

0.2 190.94% 0.08 125.27% -

0.3 139.79% 0.2 191.92% -

0.6 122.12% 0.3 139.90% -

0.8 120.33% 0.6 122.02% -

1.0 149.71% 0.8 120.32% -

- 1.0 149.56% -

Based upon these results, the CPPU/CLTP scaling factor of 1.5 was conservatively used.

Although some of the ratios exceed 1.5, those occurrences typically are of low energy

( *<0.3 Mev) and/or the activity at that energy represented a small fraction of the total source activity such that use of 1.5 is justified.

To further substantiate the value of 1.5 as conservative in this application a series of radiation shielding calculations were performed as documented in the following tabulation for conditions of shield thickness and distance and compared to the unshielded dose rates listed above.

Post- CPPU Source Accident CLTP Dose Ratio Type Time Dose Rate Rate CPPU/CLB (Days) (mR/hr) (mR/hr)

Dose Rate with 3 feet of concrete Reactor Coolant 0 3.14E+04 3.57E+04 113.68%

1 2.81E+02 3.12E+02 110.99%

30 1.85E+01 2.05E+01 110.58%

Suppression Pool 1 2.60E+O1 2.89E+01 110.94%

Reactor Steam 0 2.16E+04 2.46E+04 113.55%

60 2.1OE-03 2.38E-03 113.37%

Unshielded Dose Rate @ 132 feet Reactor Coolant 60 4.17E+03 5.OOE+03 119.77%

Comparing these results shows that including the effects of shielding and distance does not significantly affect the calculated dose rate ratios CPPU/CLTP for the unshielded cases.

NRC Ouestion 8:

Section 8.6 (page 8-9) of your submittal states that the highest estimated dose to a critical offsite location due to radiation shine from turbine building components for CPPU is approximately 4 milli-roentgen equivalent man (mrem) per year. Provide your basis for this estimate and provide your reasoning why a 500% increase in N- 16 steam activity

Enclosure to PLA-6194 Page 10 of 16 (resulting from implementation of hydrogen water chemistry at SSES 1 and 2) would not result in any increase in this estimated dose.

PPL Response:

The estimate of 4 mrem per year is based upon updating direct and skyshine radiation dose rate calculations to critical SSES offsite locations from turbine and condensate equipment to reflect CPPU operation. These calculations utilize design basis source terms (including the effect of plant operation with Hydrogen Water Chemistry) and model onsite radiation sources that contribute to offsite dose. The modeling of the turbine and condensate radiation sources include the Unit I and 2 high and low pressure turbines, moisture separators, combined intermediate valves (CIVs), 42 inch cross around piping from the moisture separators to the C1Vs, CIV to low pressure turbine piping, the vertical and horizontal high pressure turbine inlet piping and condensate storage tanks.

The results of the calculations for determining the annual doses to critical site locations from turbine and condensate sources are listed in the following tabulation and are the bases for the reported 4 mrem per year estimate.

Annual Dose Residence in Towers Club Residence in Residence in Contributors SE Sector (7) WSW Sector (12) NNW Sector (16) WSW Sector (12)

(mrem) (mrem)1 (mrem) (mrem)

Nominal Distance from site 0.5 0.4 0.6 1.3 center (miles)

Unit I & 2 Turbine Bldg. 3.46E+00 4.03E-01 1.64E+00 2.54E-03 Skyshine Dose Unit I Turbine Bldg. Direct - 4.34E-03 Dose Unit 2 Turbine Bldg. Direct 5.98E-01 Dose Condensate Storage Tank 4.40E-03 Skyshine Dose Totals 4.06E+00 4.03E-01 1.64E+00 2.54E-03 I) Total annual dose at the Towers Club is based on 500 hour0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />s/year occupancy; other locations are 8760 hour0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br />s/year.

Susquehanna implemented hydrogen water chemistry operation beginning in 1999. With HWC implementation, the N- 16 steam activity increased by approximately 500%. The HWC impact is included in the CLTP dose and CPPU offsite dose projections.

Based upon a review of environmental TLD station data before and after HWC implementation, there is a noticeable difference in exposure readings for the dosimeters located along the station's on-site perimeter fenced areas, but little or no difference in exposure data when offsite TLD locations are considered (depending on the sector from one-half mile to beyond one mile). Similarly the additional CPPU increase of approximately 14% is not expected to result in a noticeable difference in off-site exposures.

Enclosure to PLA-6194 Page 11 of 16 NRC Ouestion 9:

Discuss any effects that the storage of the higher irradiated (due to the increased core flux) spent fuel assemblies in the spent fuel pool (SFP) may have on dose rates in accessible areas adjacent to the sides or bottom of the SFP. Discuss any plans that you may have (such as shuffling of spent fuel assemblies in the SFP so that the older assemblies are located at the perimeter of the SFP) to minimize the effects of the storage of the higher irradiated spent fuel assemblies in the SFP on dose rates in areas surrounding the SFP.

PPL Response:

The design of the spent fuel pools is conservative from the radiation exposure perspective such that small changes in the fuel inventory will have inconsequential changes in operating dose rates. The SFP shielding is a minimum of 5 to 5-1/2 ft of concrete along the sides and bottom. As stated in Section 6.3.3 of the PUSAR, the normal radiation levels around the SFP with CPPU may increase by approximately 14%, primarily during fuel handling operations. Radiation zoning in accessible areas surrounding the SFP are typically zone II (< 2.5 mr/hr) or III (< 15 mr/hr). Current survey measurements taken in these areas show readings that are < 2 mr/hr. With the projected increase of 14% for CPPU, the radiation zone II and III limits surrounding the SFP will be maintained.

SSES has no plans and does not anticipate the need for shuffling spent fuel assemblies in the SFP in order to maintain acceptable dose rates in the surrounding areas. Design basis shielding calculations will be updated to substantiate the projected plant specific dose rates and radiation zone designation surrounding the SFP prior to CPPU implementation.

NRC Ouestion 10:

For each of the four vital areas listed under "Vital Missions" in Table 8-1, provide the estimated vital area post-accident dose rates, which were used to determine the vital area mission dose. Provide mark-ups of plant layout maps showing the access routes to all vital areas listed in Table 8-1.

PPL Response:

The four vital mission areas listed in Table 8-1 include:

• Post Accident Sampling Station

" Radiation Chemistry Laboratory

" SFP Cooling ESW Valve Actuation o Post Accident Vent Sampling Station

Enclosure to PLA-6194 Page 12 of 16 As noted in Table 8-1 (note 8) the Post Accident Sampling System has no associated dose rate or mission dose. The remaining three vital mission areas are addressed as follows:

Radiation Chemistry Laboratory The Radiation Chemistry Laboratory is located on elevation 676 ft of the control complex. The total maximum dose rate used to determine the vital area mission dose is conservatively taken as 5.5 mrem/hr. This radiation level is the same as the maximum level estimated in the control building at the control room elevation 729 ft. As noted in Table 8-1 (note 10), CPPU mission dose is based upon Alternative Source Term (AST) contained source terms @ 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> post LOCA for a 30 minute occupancy.

Mark-ups of plant layout maps showing the Radiation Chemistry Laboratory location are provided in Attachment 1.

SFP Cooling ESW Valve Actuation The Susquehanna specific analysis of the mission to provide ESW makeup to the SFP under LOCA conditions as provided in Table 8-1 of the PUSAR is conservatively based on adjusting the operator access dose to the fuel pool cooling heat exchanger pump room to control ESW flow in three operator steps. The previously calculated CLTP doses are adjusted via scaling factors to reflect CPPU conditions. A breakdown of the mission dose for each of the three steps is presented below.

1. Ingress/Egress dose:

= {Stairwell dose + RB 749 dose} x 2 (for ingress and egress)

= {(10.1 R/hr x 1 min/60 min/hr) + (0.5 R/hr x 1.3 min. + 7.75 R/hr x 0.175 min)/60 min/hr} x 2

= (0.168 R + 0.033 R) x 2 = 0.403 R

2. Access to Panel IC206 to check skimmer surge tank water level:

= Access to Panel IC206 + Operator Dose at IC206 Panel

= Access to Panel IC206 + 49.0 R/hr x 1 min/60 min/hr Access to Panel IC206 Invress/Eeress Dose Estimate Travel Time Dose Rate Dose (sec) (R/hr) (R) 4.3 10.5 0.0126 3.0 13.0 0.0108 3.0 18.2 0.0151 3.0 26.0 0.0216 3.0 39.0 0.0325 Total 0.0927 Ingress/Egress x 2 0.185

Enclosure to PLA-6194 Page 13 of 16

=0.185 R+ 0.817 R = 1.002 R

3. Dose in heat exchanger pump room to open valves =

= normal + contained source dose rate x stay time

= (0.04 R/hr + 0.0484 R/hr) x 2 min x I hr/60 min = 0.0029 R Total CLTP dose = 0.403 + 1.002 + 0.0029 = 1.41 R The CLTP dose of 1.41 R was increased by a factor of 1.067 to account for a suppression volume input correction in the source term adjusting this dose to 1.505 R -

CPPU dose = 1.505 R x 1.5 = 2.26 R The 1.5 factor is the CLIP to CPPU scaling factor discussed in the response to Question 8.

Mark-ups of plant layout maps showing the access route are provided in Attachment 1.

Post Accident Vent Sampling Station The Post Accident Vent Sampling System mission dose is determined based upon the estimated post LOCA dose rates and exposure times when performing the mission in accordance with the SSES Chemistry Sampling Team Emergency Plan Position Specific Instruction. This mission dose has been updated to reflect CPPU and AST implementation. A breakdown of dose rates used in arriving at the mission dose is provided in the following Table. The CPPU mission dose is conservatively increased by a factor of 1.2 to account for isotopes not included in the AST RADTRAD 60 isotope source term definition.

Mark-ups of plant layout maps showing the access routes/locations are provided in .

Post Accident Vent Sampling System Mission Doses TASK DESCRIPTION Task Whole Body Whole Extremity Extremity Duration Dose Rate Body Dose Dose Rate Dose (minutes) (R/hr) (R) (R/hr) (R)

Briefing, Assignments and Preparation Of n/a 0 0 0 0 Radiochemistry Labs INGRESS-Transit From Chem. Lab to 2 0 0 0 0 PAVSS/Surveys Perform valve lineups to establish isokinetic flow 3 0 0 0 0 through PAVSS Isolate sample flow to affected SPING. Start 2 0 0 0 0 sample flow to PAVSS panel.

Enclosure to PLA-6194 Page 14 of 16 Post Accident Vent Sampling System Mission Doses TASK DESCRIPTION Task Whole Body Whole Extremity Extremity Duration Dose Rate Body Dose Dose Rate Dose (minutes) (R/hr) (R) (R/hr) (R)

Flush sample lines for a minimum of three 6 sample line volumes Adjust Rad Sample Control Valve 0 I 3.718E-03 6.196E-05 1.372E-02 2.287E-04 023 until Radiation Sample Velocity Fl-06560A is within +/- 20% of Stack Velocity FI-06562A Flush system for a minimum of three 5 0 0 0 0 sample line volumes.

Determine optimum sample time for particulate 5

/iodine grab sample.

Determine time since shutdown and NG/l 2 2.584E-04 8.613E-06 2.584E-04 8.613E-06 Ratio. Obtain Mid-Range and High-Range noble gas concentrations. Obtain Radiation Sample Flow FI-X6560B and Stack Flow FI-X06562B.

Determine optimum sample time for 3 0 0 0 0 particulate /iodine grab sample.

Obtain noble gas grab sample. 5 Attach gas sample container to local I 7.780E-03 1.297E-04 7.523E-03 1.254E-04 connections on 0C260. Ensure stopcocks on gas container are open. Position valves as follows: 0-65-027 OPEN; 0-65-028 OPEN; 0-65-026 CLOSED Allow sample to flow through gas sample 2 0 0 0 0 container for two minutes Position valves as follows: 0-65-026 1 7.780E-03 1.297E-04 7.523E-03 1.254E-04 OPEN; 0-65-027 CLOSED; 0-65-028 CLOSED. Close stopcocks on gas container are open.

Place noble gas grab sample in plastic I 7.780E-03 1.297E-04 7.523E-03 1.254E-04 bag and seal. Obtain contact dose rate on noble gas grab sample. Transfer to sample cask.

Obtain particulate /iodine grab sample from 42 vent Panel 0C259 Valve Lineups-Valve 0 2 3.424E-03 1.141E-04 7.646E-03 2.549E-04 017 OPEN; Valve 0-65-021 CLOSED Sampling valve lineups to remove 4 4.571 E-03 3.047E-04 1.335E-02 8.899E-04 original sample cart, cart disconnects, valve lineups to install new sample cart.

Sample cart movements. Remove 12 7.144E-04 1.429E-04 7.144E-04 1.429E-04 original cart, install new cart.

Obtain particulate/iodine sample. 4 3.424E-03 2.283E-04 7.646E-03 5.097E-04 Sampling valve lineups to begin sampling, sampling valve lineups to secure grab sample.

Enclosure to PLA-6194 Page 15 of 16 Post Accident Vent Sampling System Mission Doses TASK DESCRIPTION Task Whole Body Whole Extremity Extremity Duration Dose Rate Body Dose Dose Rate Dose (minutes) (R/hr) (R) (R/hr) (R)

Sampling valve lineups to remove new 2 4.571 E-03 1.524E-04 1.335E-02 4.449E-04 sample cart with particulate/iodine sample Remove sample cart containing 6 7.144E-04 7.144E-05 7.144E-04 7.144E-05 particulate/iodine sample.

Sampling valve lineups to install original 2 4.571 E-03 1.524E-04 1.335E-02 4.449E-04 sample cart Install original sample cart 10 7.144E-04 1.191E-04 7.144E-04 1.191E-04 Obtain contact dose rate on sample cart. 11 Obtain contact dose rate 2 7.144E-04 2.381 E-05 7.144E-04 2.38 1E-05 Remove cart to low dose location 9 1.250E-06 1.875E-07 1.250E-06 1.875E-07 Measure particulate/iodine sample filter dose 12 rate Prepare to remove sample cart lid 2 1.250E-06 4.167E-08 1.250E-06 4.167E-08 Remove sample cart lid 4 9.840E-05 6.560E-06 9.840E-05 6.560E-06 Obtain filter assembly contact dose rate, 6 2.900E-04 2.900E-05 1.070E-03 1.070E-04 Release quick disconnect and remove filter assembly from cart. Separate filter assembly, remove iodine cartridge and place in plastic bag, obtain contact dose rate on iodine cartridge, transfer cartridge to sample cask and obtain contact dose rate on iodine cask. Remove particulate cartridge and place in plastic bag, obtain contact dose rate on particulate cartridge, transfer cartridge to sample cask and obtain contact dose rate on particulate cask.

Notify Chemistry Coordinator of sample and 1 8.080E-06 1.347E-07 8.080E-06 1.347E-07 cask dose rates.

EGRESS-Transit from chem. lab to PAVSS 3 8.080E-06 4.040E-07 8.080E-06 4.040E-07 Totals 1.80E-03 3.63E-03 With factor of 1.2 added 2.17E-03 4.36E-03

Enclosure to PLA-6194 Page 16 of 16 NRC Question 11:

Section 8.6 of your submittal states that the transport and storage of radioactive materials pathway is the major source of offsite dose, contributing approximately 12.2 mrem of the estimated 13.6 mrem/year to the limiting dose receptor location subject to the limits of 40 CFR 190 (25 mrem/year from effluents and external shine). Provide a breakdown for the estimated dose contributions from the other dose pathways (liquid radioactive effluents, gaseous radioactive effluents, and gamma radiation shine from the plant turbines) that make up this 13.6 mrem/year estimate.

PPL Response:

The maximum annual dose to members of the public from all contributing SSES fuel cycle components with CPPU is conservatively estimated to be 13.4 mrem total body and 13.6 mrem organ. This dose is calculated at the limiting site location (i.e., Towers Club) and is less than the 40 CFR 190 annual limits. A breakdown of the estimated dose pathway contributions to this limiting dose is provided below.

Annual Dose (mrem)

Location 2 Dose Contributor Description Towers Club WSW Sector (0.4 mi)

U I & U2 Turbine Building: Skyshine 4.03E-01 U I Turbine Building: Direct U2 Turbine Building: Direct U2 Condensate Storage Tank Skyshine Temporary Laundry Facility 3.53E-02 LLRWSF 1.14E-01 ISFSI Storage 2.32E+00 Storage of Radioactive Material at LLRWSF 5.70E+00 Transport of Radioactive Material 1.40E+00 ISFSI Transport 3.60E-02 SEALAND Container Storage 2.58E+00 Subtotal (all except effluent) 1.26E+01 Total Body Dose Calculations SSES Airborne Effluent: 7.54E-01 SSES Liquid Effluent 1.67E-02 Summed Dose: Total Body (mrem) 1.34E+01 Organ Dose Calculations SSES Airborne Effluent.: Organ 7.93E-01 SSES Liquid Effluent.: Organ 2.17E-01 Summed Dose: Organ (mrem) 1.36E+01 Notes: Annual occupancy has been applied based on 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> per year.

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