ML033280022
| ML033280022 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 11/13/2003 |
| From: | Jamil D Duke Power Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TAC MB7014, TAC MB7015 | |
| Download: ML033280022 (35) | |
Text
. \\X Ph Duke PrPowere A Duke Energy Company D.M. JAMIL Vice President Duke Power Catawba Nuclear Station 4800 Concord Rd. / CNO1 VP York, SC 29745-9635 803 831 4251 803 831 3221 fax November 13, 2003 U.S. Nuclear Regulatory Commission Attention:
Document Control Desk Washington, D.C. 20555
Subject:
Duke Energy Corporation Catawba Nuclear Station, Units 1 and 2 Docket Numbers 50-413 and 50-414 Proposed Technical Specifications and Bases Amendment Technical Specification and Bases 3.6.10 Annulus Ventilation System (AVS)
Technical Specification and Bases 3.6.16 Reactor Building Technical Specification Bases 3.7.10 Control Room Area Ventilation System (CRAVS)
Technical Specification Bases 3.7.12 Auxiliary Building Filtered Ventilation Exhaust System (ABFVES)
Technical Specification Bases 3.7.13 Fuel Handling Ventilation Exhaust System (FHVES)
Technical Specification and Bases 3.9.3 Containment Penetrations Technical Specification 5.5.11 Ventilation Filter Testing Program (VFTP)
TAC Numbers MB7014 and MB7015
References:
- 1.
Letter from Duke Energy Corporation to NRC, same subject, dated November 25, 2002
- 2.
Letter from NRC to Duke Energy Corporation, Request for Additional Information, dated September 11, 2003 Reference 1 transmitted a license amendment request concerning the subject TS and Bases sections involving ventilation systems.
The NRC transmitted a request for additional information via Reference 2.
The purpose of this Abo(
www.duke-energy. corn
x U.S. Nuclear Regulatory Commission Page 2 November 13, 2003 letter is to respond to the request for additional information. to this letter contains Duke Energy Corporation's response.
The format of Attachment 1. is to restate the NRC question, followed by Duke Energy Corporation's response.
Duke Energy Corporation has determined that the original No Significant Hazards Consideration Analysis and Environmental Analysis contained in Reference 1 are unchanged as a result of this response.
Pursuant to 10 CFR 50.91, a copy of this letter is being sent to the appropriate State of South Carolina official.
Inquiries on this matter should be directed to L.J. Rudy at (803) 831-3084.
Very truly yours, D.M. Jami Attachment
U.S. Nuclear Regulatory Commission Page 3 November 13, 2003 D.M. Jamil affirms that he is the person who subscribed his name to the foregoing statement, and that all the matters and facts set forth herein are true and correct to the best of his knowledge.
Subscribed and sworn to me:
//- 13 -,200°3 Date Nota:
My commission expires:
7-1/ -
l2 Date BE; SEAL
i U.S. Nuclear Regulatory Commission Page 4 November 13, 2003 xc (with attachment):
L.A. Reyes U.S. Nuclear Regulatory Commission Regional Administrator, Region II Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, GA 30303 E.F. Guthrie Senior Resident Inspector (CNS)
U.S. Nuclear Regulatory Commission Catawba Nuclear Station S.E. Peters (addressee only)
NRC Senior Project Manager (CNS)
U.S. Nuclear Regulatory Commission Mail Stop 08-H12 Washington, D.C. 20555-0001 H.J. Porter, Director Division of Radioactive Waste Management Bureau of Land and Waste Management Department of Health and Environmental Control 2600 Bull St.
Columbia, SC 29201
ATTACHMENT I
RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION
REQUEST FOR ADDITIONAL INFORMATION DUKE POWER COMPANY CATAWBA NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414 The Nuclear Regulatory Commission (NRC) staff has reviewed the licensee's submittal dated November 25, 2002, regarding proposed changes Technical Specifications for five ventilation systems, and the reactor building and containment penetrations.
The NRC staff has identified the following information that is needed to enable the continuation of its review.
Dose Analysis
- 1.
The application dated November 25, 2002, includes an analysis of the radiological consequences of a design basis loss of coolant accident (LOCA) using alternative source term (AST) methodology.
The application also states the NRC has previously approved the AST methodology for the fuel handling and wier gate drop accidents and that the additional use of AST methodology for the LOCA analysis would represent full implementation of the AST methodology for Catawba.
There is at least one other application for Catawba that is currently under review that may be affected by the proposed use of AST methodology in the November 25, 2002, application and that is the application dated February 27, 2003, related to the use of mixed oxide (MOX) lead test assemblies.
This issue was addressed in requests for additional information on the MOX application that were issued on July 25, 2003, that addressed how the AST related issues in the two applications would be coordinated.
Accordingly, this request for information on the November 25, 2002, application now includes the following request on how the proposed changes included in this application will be coordinated with Page 1
other pending Catawba applications, including the MOX application.
The licensee's February 27, 2003, application to allow the use of four MOX fuel assemblies at either McGuire or Catawba, currently under NRC staff review, does not address the proposed implementation of a full scope alternative source term at Catawba with total effective dose equivalent (TEDE) criteria for the design basis accident dose analyses.
The November 25, 2002, application does not address the effect of the proposed MOX lead test assemblies on the dose consequences of design basis accidents.
Please explain how the differences between these two applications will be resolved, including comparable differences for any other proposed license amendments for Catawba that are currently pending.
Duke Energy Corporation Response:
In the February 27, 2003 letter (Reference 1) Duke Energy Corporation requested a change to the licenses and Technical Specifications for McGuire Nuclear Station (McGuire) and Catawba Nuclear Station (Catawba) to allow the use of four mixed oxide (MOX) fuel assemblies in a McGuire or Catawba reactor. As noted, at the time when this submittal was made, the fuel handling and weir gate drop accidents for Catawba had incorporated the Alternative Source Term (AST) methodology as part of their licensing basis analyses.
This methodology was submitted to the NRC as a License Amendment Request on December 20, 2001 (Reference 2) and approved by the NRC on April 23, 2002 (Reference 3).
The LOCA analysis AST methodology presented in this submittal for Catawba was in the NRC review process. The McGuire radiological licensing analysis for these events was based upon earlier methodologies.
Duke Energy Corporation is continuing on a program to complete submittals to incorporate AST methodology for McGuire in early 2004.
Due to this sequence of Duke Energy Corporation submittals for Catawba and McGuire and the NRC review schedules for these submittals, the original License Amendment Request on the use of MOX fuel lead assemblies did not address the proposed implementation process for a full scope AST licensing basis for Catawba.
The process discussion was Page 2
Y complicated by the status of the McGuire analysis and the related control room testing program that is still underway.
However, as a result of additional program developments for the MOX fuel lead assembly fabrication and delivery schedule, Duke Energy Corporation has recently informed the NRC that the MOX fuel lead assembly program will be focused solely on Catawba (Reference 4).
This action couples the MOX fuel lead assembly program only to the Catawba licensing basis, which currently includes AST methodology for the fuel.
handling and weir gate drop accidents in accordance with Regulatory Guide 1.183 and TID source term for all other events in accordance with corresponding regulatory guidance.
Duke Energy Corporation intends to utilize AST methodology in support of facility operation with both low enriched uranium (LEU) and MOX fuel.
Duke Energy Corporation considers the application of AST at Catawba and the use of MOX fuel as separate licensing actions, so that the License Amendment Requests for the use of AST and for operation with MOX fuel have been made separately.
The MOX lead fuel assembly submittal evaluations and analyses are intended to be consistent with the current licensing basis for Catawba.
The details of this approach were provided in response to the Request for Additional Information (RAI) made in regard to that submittal (Reference 5).
In particular, the response addressed the similarities and differences for MOX fuel regarding the release fractions provided in Regulatory Guide 1.183 Section 3, "Accident Source Term."
The responses to this portion of that RAI were submitted on November 3, 2003.
In summary with respect to MOX fuel, the analyses related to fuel handling and weir gate drop accidents are performed in accordance with AST and reported in terms of TEDE, and all others are performed in accordance with TID source term and reported in terms of whole body and thyroid doses. When the Catawba LOCA AST License Amendment Request review is completed, Catawba will then be licensed to AST and the going-forward analyses and evaluations will be performed in accordance with that new licensing basis. Where necessary and when appropriate, for example, with respect to the LOCA accident evaluation and analyses with four MOX lead fuel assemblies, conforming changes in results related to the use of MOX fuel will be incorporated into the licensing basis.
This is the process by which the differences in these two applications will be resolved.
There are no comparable differences for any other license amendments for Catawba Page 3
that are currently pending.
It is expected that when the future License Amendment Request will be made to support MOX fuel batch operation in Catawba and McGuire, the licensing basis for Catawba will be based upon AST and the NRC review of the corresponding AST licensing submittals for McGuire will be in progress or completed.
In accordance with Regulatory Guide 1.183, after implementation of AST is complete, it is anticipated that there may be a subsequent need to revise other design basis accidents.
For these recalculations, Duke Energy Corporation intends to meet stated NRC expectations that all characteristics of the AST and the TEDE criteria incorporated into the design basis will be addressed in all affected analyses on an individual as needed basis.
In particular, this would be done where reevaluation using the previously approved source term assumptions would not be appropriate.
Accordingly, analyses of other accident scenarios are being performed in anticipation of the NRC approval of AST at Catawba as the appropriate opportunity arises.
For example, the Steam Generator Tube Rupture and the Main Steam Line Break accident analyses for Catawba are projected to be completed by the end of 2003 and are being performed in accordance with Regulatory Guide 1.183.
These future submittals will address facility operation with both LEU and MOX fuel to the extent that has been proposed to the NRC through the License Amendment Request process.
- 2.
The November 25, 2002, application proposes to use organ dose weighting factors given in International Commission on Radiological Protection Publication 60, "1990 Recommendations of the International Commission on Radiological Protection" (ICRP-60).
SECY 01-0148, "Processes for Revision of 10 CFR Part 20 Regarding Adoption of ICRP Recommendations on Occupational Dose Limits and Dosimetric Models and Parameters,"
addresses the staff position on ICRP-60.
The Commission directed the staff not to adopt ICRP-60 at that time (Staff Requirements -
SECY-01-0148, "Processes for Revision of 10 CFR Part 20 Regarding Adoption of ICRP Recommendations on Occupational Dose Limits and Dosimetric Models and Parameters"), but to monitor the work of other federal agencies and the revision to ICRP-60, which is ongoing.
The NRC staff believes that it is Page 4
premature to consider adoption of ICRP-60 at this time.
On October 17, 1994, in the Federal Register (59 FR 52255), the NRC proposed revisions to 10 CFR Parts 50, 52 and 100 (References 1, 2, and 3) that included definitions of TEDE, deep-dose equivalent, and committed effective dose equivalent.
The statements of consideration for this rulemaking noted that the definition of TEDE is meant to be consistent with 10 CFR Part 20 (Reference 4).
These definitions are currently codified in 10 CFR 50.2.
Therefore, the NRC staff believes that the organ dose weighting factors given in 10 CFR 20.1003 are the accepted values to be used in the calculation of TEDE as defined in 10 CFR 50.2.
Please provide further justification for use of the ICRP-60 organ dose weighting factors considering the regulatory definition of TEDE.
Duke Energy Corporation Response:
A synopsis of the development of the dose coefficients used in the analyses of radiological consequences of the design basis LOCA is presented in the application of November 25, 2002, Attachment 3, Appendix A, Section 1.8 (Pg. A26-A28).
There it was stated that The dose coefficients used in the analyses and the calculation of the TEDE's conform to the current regulatory positions for use of AST...with two conservative exceptions."
In our initial work, additional modeling had been incorporated to conservatively address features of dose coefficient development derived from the ICRP-60 publications (Reference 6).
In particular, conservative models were added to include a constituent for the skin in the total committed dose equivalent and to include an adjusted thyroid committed effective dose equivalent to account for the increase in the thyroid weight factor from 0.03 to 0.05 as described in ICRP-60.
This question requests further demonstration of the conservatisms identified in the application of these dose coefficients and also describes regulatory positions on current applicability of ICRP-60 organ dose weighting factors in regulatory analysis.
In response to this question, Duke Energy Corporation has performed additional reviews of the positions established by the staff and the Commission as outlined in SECY 01-0148, "Processes for Page 5
Revision of 10 CFR Part 20 Regarding Adoption of ICRP Recommendations on Occupational Dose Limits and Dosimetric Models and Parameters" (Reference 7), and the Staff Requirements Memorandum directing NRC staff actions on SECY 01-0148 (April 12, 2002) (Reference 8).
These documents demonstrate the deliberate process being followed by the NRC, along with other governmental agencies, to identify appropriate changes to 10 CFR Part 20.
Our review also focused on the Final Rule on the "Use of Alternative Source Terms at Operating Reactors" (10 CFR Parts 21, 50, and 54; RIN 3150-AG12) (Reference 9).
Key factors pertaining to the regulatory position Duke Energy Corporation needs to propose are described in Part III, "Section-by-Section Analysis," and in particular under D.
"Sections 50.67(b)(2)(i), (ii), (iii)."
Some of the same considerations and rationale that were developed to establish future LWR siting criteria are applied here (see the Final Rule on "Reactor Site Criteria Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants" (Reference 10)).
The rationale provided below is used to support the NRC position that a "capping" limitation, an additional requirement that the dose to any individual organ not be in excess of some fraction of the total as provided for routine occupational exposures, is not required.
This evaluation can be extended as part of Duke Energy Corporation's reconsideration of applying added conservatism to AST licensing basis methods for accident analysis.
-...this non-inclusion of a "capping" limitation is consistent with the final rule published in December 11, 1996 (61 FR 65157), with regard to doses to persons offsite.
Second, the use of 0.05 Sv (5 rem) TEDE as the control room criterion does not imply that this would be an acceptable exposure during emergency conditions, or that other radiation protection standards of Part 20, including individual organ dose limits, might not apply.
This criterion is provided only to assess the acceptability of design provisions for protecting control room operators under postulated DBA conditions.
The DBA conditions assumed in these analyses, although credible, generally do not represent actual accident sequences, but are specified as conservative surrogates to create bounding conditions for assessing the acceptability of engineered safety features.
- Third, 20.1206 permits a once-in-a-lifetime planned special dose of five times the annual dose limits.
Also, Page 6
Environmental Protection Agency (EPA) guidance sets a limit of five times the annual dose limits for workers performing emergency services such as lifesaving or protection of large populations.
Considering the individual organ weighting factors of
§20.1003 and assuming that only the exposure from a single organ contributed to TEDE, the organ dose, although exceeding the dose specified in §20.1201(a), would be less than that considered acceptable as a planned special dose or as an emergency worker dose.
The NRC is not suggesting that control room dose during an accident can be treated as a planned special exposure or that the EPA emergency worker dose limits are an alternative to GDC-19 or the final rule.
However, the NRC does believe that these provisions offer a useful perspective that supports the conclusion that the organ doses implied by the 0.05 Sv (5 rem) criterion can be considered to be acceptable due to the relatively low probability of the events that could result in doses of this magnitude. "
Duke Energy Corporation recognizes the similarities in criteria, but the differences in focus between controlled occupational dose evaluations in 10 CFR Part 20 and the accident analysis evaluations for 10 CFR Part 50.67 and 10 CFR 50 AppendixA, GDC-19. With regard to AST accident analyses, Duke Energy Corporation concludes that the rulemaking and resultant guidance for AST application were established with knowledge of the developments in ICRP-60.
This process and the decision not to impose an additional "capping" limitation are supported by the conservatisms in the determination of bounding conditions in DBA analyses
("conservative surrogates"), the conservatisms in dose limits (EPA guidance), and the recognition of the low probability of these events.
Therefore, consistent with staff guidance and results of rulemaking, Duke Energy Corporation will establish the Catawba AST licensing basis consistent with the guidance presented in Regulatory Guide 1.183, the final AST Rule, the AST Standard Review Plan, and 10 CFR 50.67 and 10 CFR 50 Appendix A, GDC-19. Added conservatism in determination of dose coefficients or "capping" limitation approaches will not be required in the evaluation of postulated DBA conditions using AST.
Duke Energy Corporation will strive to complete revisions to the supporting analyses for this Page 7
license amendment request and submit these result revisions to the NRC by November 30, 2003.
- 3.
The November 25, 2002, application proposes to eliminate the annulus ventilation system (AVS) one minute drawdown time surveillance requirement (SR) in SR 3.6.16.2.
In justification of this change, it is stated on page 5 of Attachment 3 of the November 25, 2002, application that the one minute annulus drawdown time is not a dose analysis input.
In the LOCA dose consequences analysis, provided in Appendix A of the submittal, the leakage from the containment is assumed to be an unfiltered ground level release prior to AVS drawdown.
After 23 seconds, the release is assumed to be filtered by the AVS and released from the plant vent.
This seems to contradict the claim that the drawdown time is not a dose analysis input.
Please provide further information on the treatment of containment leakage by the AVS, including the time it takes for the AVS to drawdown the annulus to the required negative pressure.
Additionally, please provide further justification on why a surveillance is not needed to verify that the AVS can provide filtration of the containment leakage as assumed in your dose analyses.
Duke Energy Corporation Response:
The submittal of November 25, 2002 lists the response time of the AVS as 23 seconds (Pg. A-50 and A-73).
This is the upper bound of the time required for the AVS to activate and for its fans to come to full speed.
This equipment operation time is an input to the dose analysis model.
In addition, an analysis has been performed to determine that the AVS draws the annulus to the required negative pressure within at most 41.4 seconds after the initiating event for a design basis LOCA with Minimum Safeguards (only one AVS fan in operation -
cf. Pg. A-52).
The analysis also has shown that the AVS draws the annulus to the required negative pressure within at most 30.5 sec for all other design basis LOCA scenarios (two AVS fans in operation -
cf. Pg. A-53 and A-54).
Only at these times is credit taken for filtration of the containment leakage into the annulus by the AVS. The analysis also shows that the AVS draws the annulus pressure to the setpoint for modulation between Exhaust and Page 8
Recirculation at 54 sec for the design basis LOCA with Minimum Safeguards and 34 sec for the other design basis LOCA scenarios (Pg. A A-54).
It is at this point that postulated failure of one AVS pressure transmitter manifests itself and causes its AVS fan to continue in full exhaust for the design basis LOCA with an AVS pressure transmitter failure.
As noted above, all these drawdown times are calculated values, including the times for drawdown of the annulus to the required negative pressure.
The calculation of time of drawdown of the annulus by the AVS takes into consideration the following variables:
3.1)
Rate of inleakage across the reactor building into the annulus.
3.2)
AVS fan flow rate.
3.3)
AVS response time (defined above as 23 seconds).
3.4)
Rate of containment leakage into the annulus.
3.5)
Limiting temperature and pressure in containment.
3.6)
Mass of air in the annulus.
3.7)
Limiting temperature and pressure for the outside air.
3.8)
Limiting initial temperature and pressure in the annulus.
3.9)
Limiting difference in hydrostatic gradients of the air in the annulus and the outside air (taking appropriately limiting conditions for the air in the annulus and the outside air).
3.10) Rate of heat-up of the containment shell, convective heat transfer from the containment into the annulus, rate of thermal expansion of the containment shell into the annulus.
3.11) Pressure inside the containment and rate of pressure induced expansion of the containment shell into the annulus.
3.12) Accumulation of water vapor into the annulus and rate of direct radiation heat transfer from the containment shell to the annulus atmosphere.
No tests or surveillances are associated with the latter six sets of variables. Limiting conditions associated with these twelve variables are used in the analysis.
The mass of the air in the annulus is fixed by the initial conditions (pressure, temperature) taken for the annulus and the size (volume) of the annulus. Technical Specification (TS) 5.5.2 gives the acceptance criterion for the test for rate of Page 9
containment leakage.
TS 3.6.4 and 3.6.5 give the limits for containment temperature and pressure.
The tests of AVS response time are conducted with a test acceptance criterion of 23 seconds.
Surveillance Requirement (SR) 3.6.10.5 gives the acceptance criterion for the flow rate for each AVS fan.
This leaves the rate of inleakage across the reactor building to be determined. A limiting value, 2000 scfm at -
1 in.w.g., is taken in the calculations of post LOCA conditions in the annulus. A mass rate of inleakage is calculated taking the 95th percentile low for outside air temperature.
Now it remains to verify that reactor building inleakage does not exceed 2000 scfm at an annulus pressure of -1 in.w.g.
The current test procedure used to determine the rate of reactor building inleakage is called the drawdown test.
The current acceptance criterion for the drawdown test, cited in SR 3.6.16.2, is that the AVS shall draw the annulus to -0.5 in.w.g. within 1 minute after a start signal.
It has been determined by Duke Energy Corporation that this acceptance criterion does not confirm that the rate of reactor building inleakage remains below the limiting value taken in the analysis of post LOCA conditions in the annulus (again, 2000 scfm at -1 in.w.g.).
As discussed in the November 25, 2002 submittal (Attachment 3, page 3), the required test conditions and timing for a drawdown would be very short and the measurement uncertainties with this testing technique can prevent meaningful comparisons.
The acceptance criterion is developed considering rate of reactor building inleakage, AVS fan flow rate, size of the annulus and initial conditions within, and time for the AVS fans to reach full speed and rated flow (a constituent of the AVS response time).
Duke Energy Corporation calculated a revised criterion for this test method.
The acceptance criterion requires an annulus pressure of at most -1.19 in.w.g. within 16 seconds after AVS fan start.
This testing is difficult to conduct and achieve results with the required low degree of uncertainty, as discussed in the November 25, 2002 submittal (Attachment 3, page 3).
This is the reason for the proposed change in testing methodology and acceptance criteria as described below.
Duke Energy Corporation has also conducted vacuum decay tests.
As noted in the submittal (Attachment 3, page 4),
the vacuum decay test measures the time taken for the pressure in the annulus to increase or "decay" from -3.5 Page 10
in.w.g. to -0.5 in.w.g.
This acceptance criterion (87 seconds) is determined considering the rate of reactor building inleakage and the size and initial conditions of the annulus. By following this process, the vacuum decay test provides a method for validating that the rate of reactor building inleakage does not exceed the limit used in the calculations of post LOCA conditions in the annulus.
A summary of this method of evaluation is provided as follows: The variables affecting the calculation of post LOCA annulus drawdown time are known. Acceptance criteria are given for some of these variables.
Substituting the vacuum decay test for the annulus drawdown test does not decrease the number of acceptance criteria for these variables.
It does not change the list of variables for which acceptance criteria must be met, and the number of variables with no associated acceptance criteria does not increase.
The vacuum decay test provides a more reliable method for validating the rate of reactor building inleakage than the annulus drawdown test. With the current acceptance criteria noted above, the vacuum decay test in place of the drawdown test will suffice to demonstrate the required performance of the reactor building and AVS.
Control Room Relative Concentration (X/Q)
Estimates
- 4.
What are the release heights and distances between the postulated release location and receptor? Are distances straight line or do they factor in flow over or around structures? For example, does "900 arc" mean that the release is assumed to occur due east of the receptor, but is assumed to initially move in an arc around a structure to get to the receptor?
Duke Energy Corporation Response:
Release heights and distances between release locations are provided in a separate enclosure on the following page.
Distances are straight line, unless they are denoted as "arc."
If the term "arc" is used with the distance or wind direction, then it was assumed that the effluent would have to move in an arc around the structure to reach the receptor (Control Room Area Ventilation System -
CRAVS -
outside air intake).
All arc lengths used are the shortest horizontal distances around the structure to the receptor. Vertical Page 11
1=5=5 Tabl 7 MqnSource Parameters fr ARrON96 Modelingy Icev.
¢ Sorce Type:
EQ FUEL NDOG VNdog ODOG VOdog RX FVST AGin AGout UV UVF VP VerctalPoint IR%
i M.
IN x
x x
Hotton (a!or I-1CR2 fl
-R
-R CappedPoint 8 2-CRI 2-CRI 2-CRI X
X X
x Horfoflta Area
% A x m VerficalArea C
-CRI I l.CRI
[-CRl Source D 2-CR2
___M_
2-CR2 2-CR2 X
Release Pt,6.32 m 12.5 m 12.3 m Height dID Om 20 m 12.5 m O m 12.3 m O m O m 14.6 m 6.6 m 6.6 m 38 m 38m 5.33 n Flow Rate (m3/s)
O O
O O
0 0
0 0
0 0
2.83 15.64 0
SigmaY Om 3.2m 1.3m 1.3m 1.6m Om 3.5m Om Om O
O 0
Om Om l m 1.6 m m
Sigmn-Z Om 3.3m 2.1m Im 2.1m Om 6.9m Om Om Om O
mj Om Om 1 m I m I
Bldg Cross-opposite opposite opposite opposite sectional Area 1592 m2 1592 1592 1592m 2 unit:
unit:
1592 unit:
1592 unit:
1592 1592 1592 (RXorODOG) m2 m2 1592 m2 1592 m2 m2 1592 m2 m2 1592 m2 mr 2
m 2 m 2 same same same same unit:
unit:
unit:
unit:
188.1 188.1 rn2 188.1 m2 188.1 m2 m2 Source/Stack 0
0 0
0 0
0 0
0 0
0 0
0 0
Radius* (m)
I Vertical Vel.*
0 m/s O m/s Oi ns O m/s O m/s Om/s Om/s Om/s Om/s Om/s O
Om/s I O m/s Om/s I-CRI 46 mac 80m 40 marc 40 mac 11 m I I m 1.8 m 52 m 44 m3 c loim 43 m 43 m 61.2m,8 DistanceWD 90 are 830 360°arc 360°arc 1360 1360 420 1360 3600 120 530 530 800arc arc I-CR2 125rm 128rm 92rm 92m 135m 135m 94m 160m 83m 155m 109m r
109m r
95.4rm DistanceWD 1650 1420 1780 1780 1800 1800 1740 1690 1780 20 1630 1630 1650 2-CRI 125 m 128 m 92 m 92 m 135 m 135 m 94 m 170 m 83 m 155 m 109 ml 109 m 95.4 m DistanceWD 170 400 40 40 20 20 70 140 40 1810 200 200 170 2-CR2 46m.,
80rm 40marc 40rarc lm Im 1.8m 52 m 44marc lOm 43m 43m 61.2ma.
Distance.WD 9o arc 970 180° arc 180° arc 470 470 1350 470 180° arc 1700 l
1290 129 98°arc 1:=>:*
Values of zero are assumed for the vertical velocity and stack radius parameters, in order to treat the release as a ground-level release in ARCON96. '=<=
arc lengths were not used as input; if the effluent was assumed to go over a structure, then the straight line distance was used, as if the structure was not there.
For some release locations, the release was treated as both a
4.1) horizontal/capped point release for source and receptor located on the same unit and 4.2) vertical area source release for source and receptor located on opposite units (at larger distances).
In this second case, the release height for the source was set to 0.
These release locations include the equipment hatch (denoted as EQ), and the inboard and outboard steam generator doghouses (denoted as VNdog and VOdog).
- 5.
Provide a figure or figures showing structures, assumed paths of air flow, dimensions, heights and distances used as input in estimating the postulated transport of effluent from each of the release locations to the receptors.
Are all directional inputs defined in terms of true north?
If the figures are drawn to plant or magnetic north, what is the relationship to true north, assuming that the meteorological measurements are based upon true north?
Duke Energy Corporation Response:
Figure 1 on the next page provides a plot plan of nearby structures with a scale, true north, and the plant's "Called North" orientation. As denoted on the figure, True North lies 1 11' 05" west of "Called North."
Wind directions are measured and input in terms of True North.
- 6.
If more than one release to the environment and transport scenario could occur (e.g.,
loss of offsite power, availability of offsite power, single failure), were comparative X/Q calculations made to ensure consideration of the limiting dose? Page 12
4.°I1' 05" CALLED NORTH TRUE NORTH EXTERIOR' DOGHOUSE EXTERIOR DOGHOUSE SCALE 1-100
.UVK PUW R CATAWBA NUCLEAR STATION PLOT PLAN FOR CONTROL ROOM HABITABILITY ASSESSMENT FIGURE: 1
Duke Energy Corporation Response:
The release-intake pairs were identified from a review of plant arrangement drawings and other design documentation, as well as several walkdowns inside and outside the plant (cf. Pg. A-20 of the submittal of November 25, 2002).
It was determined that the release-intake pairs did not vary with availability of offsite power (offsite power available or lost) or with any single failure with the following exceptions.
6.1)
One exception is associated with boiloff from the steam generators (S/Gs) following reactor trip.
The design basis control room X/Qs for S/G boiloff is associated with loss of offsite power. For a design basis accident with loss of offsite power, the boiloff from the S/Gs would be released through the SIG power operated relief valves (PORVs) and the Main Steam Code Safety Valves (MSSVs).
Should offsite power remain available, steam from the S/Gs would be routed through the Main Steam Bypass to Condenser System valves to the condenser.
From there, radioactive effluents would be drawn through the Condensate Steam Air Ejectors and be released through the unit vent stack to the environment.
Therefore, the control room X/Q for SIG releases for an accident with offsite power available would be associated with releases from the unit vent stack.
This control room X/Q would be represented by lower values than the control room X/Q for releases from the SIG doghouses.
Therefore, the control room X/Q for releases from the SIG doghouses is used for analyses of radiological consequences of design basis accidents involving SIG boiloff.
6.2)
The affected unit may remain on line for some time following a design basis Steam Generator Tube Rupture (SGTR).
Therefore, SGTR break flow and SIG boiloff would be routed to the turbine and condenser.
From there, it would be released through the Condensate Steam Air Ejectors and the unit vent stack as noted above.
The control room X/Q for releases from the unit vent stack is used for the design basis SGTR for the time between the initial S/G tube break and unit trip.
The design basis SGTR includes loss of offsite power at unit trip.
As a conservative measure, after simulated unit trip following a design basis SGTR, Page 13
the control room X/Q is associated with releases from the S/G doghouse for the reasons noted above ( 6.1).
- 7.
Confirm that each of the control room intakes meets the applicable qualifications to support a credit for reduction in the X/Q values.
These qualifications include, but are not limited to, the single failure criterion for active components and the seismic and missile protection criteria.
If both control room air intakes were previously approved by the NRC staff as meeting all of the applicable qualifications and that status has not changed, the licensee's response may reference the document approving the intakes.
Also, provide the assumed flow rates for each intake used in the composite X/Q calculations.
Duke Energy Corporation Response:
As noted in the submittal (cf. Pg. A-21), Catawba Nuclear Station is equipped with two CRAVS outside air intakes.
Each intake is equipped with two safety related isolation valves in series. Normally open, these valves are provided with Class E motors and therefore are "fail as is."
There is no interface with any control which would cause any of these isolation valves to close automatically.
There is no valid single failure in the plant design basis which would place Catawba in a "single intake" configuration.
The CRAVS outside air intakes are Seismic Category I. They are located so as to protect them from tornado missiles.
The design basis value taken for the split in airflow to the CRAVS outside air intakes is 60/40 (Pg. A-74).
- 8.
Page A-36 of Appendix A to Attachment 3 of the November 25, 2002, submittal lists postulated design basis events.
Page A-68 provides some of the input information as a function of release and receptor location.
Page A-70 provides X/Q values for two of the four locations listed in a table on Page A-
- 68.
What are the pairings of postulated design basis events, release and intake locations, and X/Q values? Page 14
Duke Energy Corporation Response:
Table 8-1 lists the limiting values of the control room X/Qs for all design basis accidents for Catawba Nuclear Station.
The enclosure contained in the response to Question 4 presents the input associated with these control room X/Qs except the one associated with releases from the S/G relief valve vents.
The data associated with the control room X/Q for releases from the S/G relief valve vents are presented in Table 8-2.
To date, calculations of radiation doses (including radiation doses in the control room) have been completed in conformance to Regulatory Guide 1.183 only for the following design basis accidents.
Loss of Coolant Accident, Fuel Handling Accident in Containment, Fuel Handling Accident in the Spent Fuel Pool, and Weir Gate Drop.
Calculations of offsite and control room radiation doses following the following design basis accidents are in preparation in conformance to Regulatory Guide 1.183:
Main Steam Line Break and Steam Generator Tube Rupture.
These calculations support a license amendment request to be sent to the Staff to request removal of the License Conditions of Amendments 159/151 to Facility Operating Licenses NPF-35 and NPF-52 (Reference 11).
Radiological consequences of the remaining design basis accidents with postulated release of radioactivity will be analyzed in conformance to Regulatory Guide 1.183 as warranted by future applications.
These include the following design basis accidents:
Locked Rotor Accident, Rod Ejection Accident, and Break of a Small Line Carrying Reactor Coolant Outside Containment.
The control room X/Qs listed below were calculated in conformance to Regulatory Guide 1.194 (Reference 12) with Page 15
one exception.
The exception is associated with the control room X/Q for releases from the vents of the S/G PORVs and MSSVs.
This control room X/Q is not associated with the design basis LOCA.
It was not used in analysis of radiological consequences of the design basis LOCA completed in support of this license amendment request. The exception is discussed as follows:
A control room X/Q was calculated for releases from the vents of the S/G PORVs and MSSVs located on the roofs of the S/G doghouses.
There are two SG doghouses associated with each nuclear unit at Catawba Nuclear Station: an outboard S/G doghouse and an inboard S/G doghouse.
Each of the two CRAVS outside air intakes is located in a corner between the containment and outboard SIG doghouse of the associated nuclear unit.
The design basis release point for the control room X/Q for releases from the SIG PORV and MSSV vents was the exhaust vent for the turbine of the turbine driven auxiliary feedwater (AFW) pump.
This vent is located on the roof of the inboard SIG doghouse only.
Of the vents on the roof of the inboard S/G doghouse, the exhaust vent for the turbine driven AFW pump is closest to the CRAVS outside air intake of the associated nuclear unit.
The ARCON96 computer code allows releases of this type to be modeled as either a ground release or a vent release.
The ground release option includes the assumption that the receptor is on the axis of the effluent plume.
This assumption ignores the effect of vertical separation. The vents of the MSSVs and S/G PORVs in the outboard S/G doghouses are 50'6" above the CRAVS intakes.
They are oriented vertically.
Steam will be released from the SIG PORVs and MSSVs straight up (nearly in the opposite direction from the nearer CRAVS outside air intake since the horizontal separation is not large) with high plume rise velocity.
The high temperature of the release will cause the exhaust plume to rise further.
This precludes any significant contamination of the nearer CRAVS outside air intake with releases from the SIG PORVs and MSSVs of the outboard SIG doghouse.
The vent at the inboard SIG doghouse closest to the nearer CRAVS outside air intake is the exhaust vent for the turbine driven AFW pump.
Contaminated effluent from the S/G PORVs and MSSVs of the outboard SIG doghouses will not reach the CRAVS outside air intake at significant levels, as noted above. For these reasons, the limiting values of control Page 16
room XQ for releases from the S/G PORVs, MSSVs, and exhaust vent of the turbine driven AFW pump should be associated with releases from the vent of the turbine driven AFW pump.
The Staff has recently released guidelines for taking plume rise into account in the calculation of control room X/Qs.
One position taken by the Staff concerning accounting for plume rise is as follows (Reference 12 Section 4):
".-the ground level X/Q calculated with ARCON96 (on the basis of physical height of the release point) may be reduced by a factor of 5...
only if
- 1) the release point is uncapped and vertically oriented and
- 2) the time dependent vertical velocity exceeds the 95t'-
percentile wind speed by a factor of 5."
Both of these conditions apply to the relief valve vents on the S/G doghouses.
The vents are uncapped and vertically oriented. The vertical velocity of the flow through these vents is greater than 5 times the 95t-percentile wind speed at the Catawba site (12 mph).
Therefore, the value of the control room X/Q for the effluent through the relief valve vents from the outboard S/G doghouse may be divided by 5.
The resulting value is calculated to be less than the control room X/Q for the vent of the turbine driven AFW pump.
As described above, this control room X/Q (for releases from the S/G PORVs, MSSVs, and turbine driven AFW pump exhaust vent) was not used in the analysis of radiological consequences of the design basis LOCA.
The control room X/Qs considered for use in the analysis of radiological consequences of the design basis LOCA are associated with the following release locations.
8.1)
Unit Vent Stack, 8.2)
Equipment Hatch, 8.3)
Containment Purge Ventilation Supply Intake Vent, and 8.4)
Refueling Water Storage Tank (RWST). Page 17
Table 8-1 Atmospheric Dispersion Factors for Transport of Radioactivity to an Outside Air Intake of the Control Room Area Ventilation System Following a Design Basis Accident (DBA) at Catawba Nuclear Station (1)
Time Span CR X/Q Associated DBA Scenarios (hr -
(2),
Sec/m3 (3))
Releases from the Unit Vent Stack Associated DBAs include the following:
0-2 1.74x10-3
- 1) Break of a Small Line Carrying Reactor 2-8 1.47xl0-Coolant Outside Containment, 8-10 6.90x10-'
- 2) Steam Generator Tube Rupture (before unit 10-24 6.14x10-'
trip),
24-96 4.45x10-'
- 3) Loss of Coolant Accident,96-720 3.11x10'
- 4) Fuel Handling Accident in Containment,
- 5) Fuel Handling Accident in the Fuel Building, and
- 6) Weir Gate Drop.
Releases from the Equipment Hatch Associated DBAs include the following:
0-2 1.59x10-3
- 1) Loss of Coolant Accident (containment 2-8 6.44x10' bypass leakage), and 8-10 3.46x10-4
- 2) Fuel Handling Accident in Containment 10-24 3.15x10' (with non recently irradiated fuel).
24-96 2.08x10-'96-720 1.24x10' Releases from the Containment Purge Ventilation Supply Vent Associated DBAs include the following:
0-2 l.llx10o-
- 1) Rod Ejection Accident (source term release 2-8 4.65x10' to containment and containment bypass 8-10 3.21x10-'
leakage), and 10-24 2.75xl0-
- 2) Loss of Coolant Accident (containment 24-96 2.18x10-'
bypass leak path).96-720 1.24x10' Releases from the Refueling Water Storage Tank (RWST)
Associated DBAs include the following:
0-2 l.92xl0-
- 1) Rod Ejection Accident (source term release 2-8 1.48x10-3 to containment and engineered safety 8-10 7.40x10-'
features -
ESF -
intersystems leakage to 10-24 6.21x10' the RWST), and 24-96 4.26x10'
- 2) Loss of Coolant Accident (ESF backleakage 96-720 2.74x10-'
to the RWST).
Releases from the Fuel Building This control room XQ is associated with a 0-2 4.41x10-'
Fuel Handling Accident in the Spent Fuel Pool 2-8 1.69x10-'
(involving non recently irradiated fuel).
8-10 9.70x10 10-24 9.59x10-5 24-96 6.75x10-96-720 4.57x10-5 Page 18
Releases from Within the Outboard SIG Doghouse (5)
This control room X/Q is associated with the 0-2 l.93xl0D-design basis Main Steam Line Break Outside 2-8 1.16x10-a Containment Upstream of the Main Steam 8-10 5.8Ox10-m Isolation Valve - MSIV.
(This places the 10-24 5.21x10-3 break in a S/G doghouse.)
24-96 3.58x10-96-720 2.35xl0-l Releases from the Yard Near the Outboard S/G Doghouse (5)
This control room X/Q is associated with the 0-2 4.20x10-2 design basis Main Steam Line Break in the Yard 2-8 3.59xlO-with Failure of the MSIV Associated with the 8-10 1.84x10-Faulted SG (and upstream of the break).
10-24 l9.70x10-3 24-96 7.89xl0 96-720 15.67x10-3 I~ ~
I Releases from the S/G PORVs, MSSVs and AFW TDP Exhaust Vent (4)
Associated DBAs include the following:
0-2 3.78x10-3
- 1)
Main Steam Line Break (intact SGs),
2-8 2.90x10-
- 2) Locked Rotor Accident, 8-10 1.70xl0-3
- 3) Rod Ejection Accident, and 10-24 1.45x10-
- 4) Steam Generator Tube Rupture (after unit 24-96 l.l9x10=-
trip and only for intact SGs).
96-72 0
9.57x10-'
Notes on Table 8-1
- 1) The set of values listed for each control room X/Q are limiting within each of the time intervals. There are two complete nuclear units at Catawba Nuclear Station. Catawba is also equipped with two Control Room Area Ventilation System (CRAVS) outside air intakes.
Therefore, for each control room X/Q, there are four combinations of release locations and receptors.
For each time span, the highest values for each of the release location -
receptor combination has been listed in Table 8-1.
- 2) The 0-2 hr control room XQs are used for the 2 hr time span of highest predicted release of radioactivity to the environment.
The 2-8 hr control room X/Qs are used for the remainder of the 0-8 hr time span.
- 3) The time periods over which control room X/Qs are defined include 0-2 hr, 2-8 hr (cf. Note 2 above), 8-24 hr, 24-96 hr, and 96-720 hr.
The 8-24 hr time span has been partitioned into 8-10 hr and 10-24 hr.
(Cf. Section A-21, A-22 and Ref. 24 of the submittal of November 25, 2002).
- 4) The design basis values for the control room X/Q for releases from the SIG PORVs, MSSVs, and turbine driven AFW pump exhaust vent are associated with the turbine driven AFW pump exhaust vent. This is located on the roof of the inboard SIG doghouse.
(Cf. Pg. 12 & 13).
- 5) The design basis values for these control room X/Qs are associated with the outboard SG doghouses.
Page 19
Table 8-2 Input Associated with Control Room X/Qs for Releases from the SIG Relief Valve Vents Input Parameter Unit 1 AFW Turbine Unit 2 AFW Turbine Exhaust to Unit 1 CR Exhaust to Unit 2 CR Intake Intake Number of 3
3 meteorological data files Height of upper wind 10 10 instrument (meter)
Height of lower wind 40 40 instrument (meter)
Units of wind speed mph mph data Type of release Ground Ground Release height 16.8 16.8 (meters)
Building cross 1571 1571 sectional area (square meters)
Effluent vertical 0
0 velocity (meter/sec)
Vent flow rate (cubic 11.0 11.0 meters/s ec)
Vent radius (meters) 0 0
Direction from intake 10.8 169.2 to source (degrees)
Distance to intake 38.2 38.2 (meters)
Intake height (meters) 1.4 1.4 Terrain level 0.0 0.0 difference (meters)
Minimum wind speed 0.5 0.5 (meters/second)
Surface roughness 0.1 0.1 length (meters)
Sector averaging 4.0 4.0 constant Wind direction sector 90 90 width (degrees)
Page 20
Bypass Leakage
- 9.
In the submittal, the licensee requested a change for penetration and bypass leakage from < 0.05 percent to < 1.0 percent for annulus ventilation, fuel handling ventilation, and auxiliary building ventilation systems.
The staff position on this issue is as outlined in Regulatory Position C.5.c and C.5.d of Regulatory Guide (RG) 1.52, Rev. 2, dated March 1978 (Reference 5).
Although, a different position was outlined in Generic Letter 83-13 (Reference 6), that position is not the current NRC staff's position.
The current staff's position is as indicated above and as outlined in Regulatory Position C.6 of RG 1.52 Rev. 3, dated June 2001.
Therefore, the penetration and bypass leakage should be <
0.05 percent for the systems discussed above.
Duke Energy Corporation Response:
Catawba is requesting a consistent application of the 1%
penetration and system bypass leakage design basis limit for economic reasons.
This request will also remove unnecessary restrictions from the Ventilation Filter Testing Program as described in TS 5.5.11.
Catawba is licensed to Regulatory Guide 1.52, Revision 2, and performs laboratory testing of carbon samples in accordance with ASTM D-3803-1989 at 302C and 95% relative humidity and Generic Letter 99-02.
With the requested change, Catawba will meet all applicable Regulatory Guide guidelines and remain conservative with respect to the organic methyl iodine penetration safety factor referenced in Regulatory Guide 1.52, Revision 3.
The 1% HEPA and carbon filters penetration and system bypass leakage limit has always been the limiting operating condition for the Unit 1 annulus ventilation, fuel handling ventilation, and auxiliary building ventilation filtration systems.
The 0.05% HEPA and carbon filters penetration and system bypass leakage limit for the Unit 2 annulus ventilation, fuel handling ventilation, and auxiliary building ventilation filtration systems often results in premature carbon filter media replacement even though the samples meet the ASTM D-3803-1989 laboratory low temperature (300C) and high humidity (95%) test acceptance criteria.
The 0.05% penetration and system bypass leakage limit Page 21
applies to a 99% efficient filtration system and imposes an unnecessary restrictive operating margin for these 95%
filtration systems.
Therefore, Catawba is requesting the Unit 2 change to be consistent with Unit 1.
As stated in Attachment 3, pages 13 and 14 of the November 25, 2002 submittal, Regulatory Guide 1.52, Revision 2, Subsection C.5.c initially linked the 0.05% penetration to a 99% efficient HEPA filter.
In Generic Letter 83-13, the NRC stated that the standard technical specifications for all power reactors did not clearly reflect the required testing requirements of HEPA filters and charcoal units and the NRC Staff assumptions used in the safety evaluations for the ESF atmospheric cleanup systems.
Generic Letter 83-13 clarified Regulatory Positions C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, by issuing revised standard technical specification surveillance requirements for the testing of HEPA filters and charcoal adsorber units.
The revised surveillance requirements clearly stated that the 0.05%
penetration and bypass leakage value is applicable when a HEPA filter or charcoal adsorber efficiency of 99% is assumed.
The revised surveillance requirements also clearly stated that the 1.0% penetration and bypass leakage value is applicable when a HEPA filter or charcoal adsorber efficiency of 95% is assumed.
In Regulatory Guide 1.52, Revision 3, dated June 2001, the following guidance from paragraph 6.3 mirrors the guidance provided in Regulatory Guide 1.52, Revision 2, and is consistent with the guidance provided in Generic Letter 83-13.
"To be credited with a 99% removal efficiency for particulate matter in accident dose evaluations, a HEPA filter bank in an ESF atmosphere cleanup system should demonstrate an aerosol leak test result of less than 0.05%
of the challenge aerosol at rated flow 10%."
Although some NRC reviewers have stated that they disagree with the position on penetration and system bypass leakage described in Generic Letter 83-13, the generic letter has never been amended or retracted.
Therefore, all current regulatory guidance indicates that a 0.05% penetration and system bypass leakage limit only applies to a 99% efficient filtration system. Page 22
e I
The following table and analysis supports Catawba's position that a 1% penetration and system bypass leakage allowance is only applicable to a 95% efficient filtration unit.
The table compares the filtration unit efficiency and penetration and system bypass leakage for the Catawba Unit 1 and Unit 2 safety related filtration units.
The Unit 1 annulus ventilation, auxiliary building ventilation, and fuel handling ventilation filtration systems design basis requirements are identical to Unit 2, except for the HEPA and carbon filter penetration and system bypass limits.
Control Unit 1 Annulus, Unit 2 Annulus, Room Auxiliary Building, Auxiliary Building, Ventilation and Fuel Handling and Fuel Handling Systems Ventilation Ventilation Filtration Systems Filtration Systems Design Filtration 99%
95%
95%
Unit Efficiency Carbon
< O.95%
< 4.0%
< 4.0%
Decontamination Efficiency HEPA and Carbon
< 0.05%
< 1.0%
< 0.05%
Filter Penetration and System Bypass Total Filtration 100%
100%
99.05%
Efficiency As shown in the table, the arbitrary application of a 0.05%
penetration and system bypass leakage limit to a 95%
filtration unit imposes an unnecessarily restrictive 0.95%
margin (100 -
99.05 = 0.95%).
If the position is taken that the extra 0.95% margin provides a "safety factor" for a 95%
filtration unit, then the question can be raised as to why a similar margin is not applied to a 99% efficient filtration unit.
I The carbon filter efficiency should be defined as the iodine that penetrates the carbon due to the actual iodine removal efficiency (determined by laboratory tests) plus the amount that penetrates and bypasses the carbon filter due to bypass leakage paths as shown by the following analysis: Page 23
is I
4 100% -
% Penetration and System Bypass -
% Carbon Decontamination Efficiency = Design Filtration Unit Efficiency For the 99% filtration units:
100% -
0.05% (bypass) -
0.95% (carbon efficiency) = 99%
For the 95% filtration units:
100% -
1% (bypass) -
4% (carbon efficiency) = 95%
If a 0.05% carbon penetration and system bypass limit is applied to a 95% filtration unit, additional margin is introduced:
100% -
0.05% (bypass) -
4% (carbon efficiency) = 95.95%
Applying a 0.05% penetration and system bypass leakage allowance to a 95% efficient-filtration unit appears to be an arbitrary assumption that does not correspond to design basis protection. Since the purpose of the TS is to provide design basis protection, it is reasonable to use a 1%
penetration and system bypass leakage allowance for a 95%
efficient filter unit.
As stated in Attachment 3, pages 14 and 15 of the November 25, 2002 submittal, Catawba assumes a particulate and elemental iodine efficiency of 95% and an organic iodine efficiency of 80% in the accident dose analyses.
The 1%
penetration and system bypass leakage limit for organic iodine leakage provides a safety factor of 4.75 and is conservative with respect to the current safety factor of 2 required by Generic Letter 99-02 and Regulatory Guide 1.52, Revision 3. The organic iodine safety factor is expected to be even larger than 4.75 because Catawba does not take credit in the dose analyses for humidity control, even though the filter unit heaters are maintained operable in accordance with the TS.
This organic iodine safety factor of 4.75 is consistent with the analysis presented in the NRC Safety Evaluation for License Amendments 90 to NPF-35 and 84 to NPF-52 for Catawba Units 1 and 2, respectively, as well as with the current design basis LOCA analysis.
Overall, the 95% filtration removal efficiency meets the design basis requirements for the annulus ventilation, fuel handling ventilation, and auxiliary building ventilation Page 24
O '
t filtration systems and provides adequate margin for the already conservative accident dose evaluations.
The 0.05%
penetration and system bypass leakage limit applies to a 99%
efficient filtration system and imposes an unnecessary restrictive operating margin for these 95% filtration systems.
Therefore, the design basis requirements for a 99%
filtration system should not be imposed upon the subject 95%
filtration systems.
Based on dose analysis assumptions and filter testing criteria used at Catawba, changing the penetration and system bypass to 1% does not create any safety concerns.
This change will provide additional testing margin for Catawba and not contradict any published regulatory guidance.
The historical and current organic methyl iodine filtration safety factor will be maintained and all licensing requirements associated with the filter units will be met.
REFERENCES (NRC REQUEST FOR ADDITIONAL INFORMATION)
- 1.
Title 10 of Code of Federal Regulations (10 CFR) Part 50, "Domestic Licensing of Production and Utilization Facilities."
- 2.
10 CFR Part 52, "Early Site Permits; Standard Design Certifications; and Combined Licenses for Nuclear Power Plants."
- 3.
10 CFR Part 100, "Reactor Site Criteria."
- 4.
10 CFR Part 20, "Standards for Protection Against Radiation."
- 5.
Regulatory Guide 1.52, "Design, Inspection, and Testing Criteria for Air Filtration and Adsorption Units of Post-Accident Engineered Safety Feature Atmosphere Cleanup Systems in Light Water Cooled Nuclear Power Plants."
- 6.
Generic Letter 83-13, "Clarification of Surveillance Requirements for HEPA Filters and Charcoal Absorbers Units in Standard Technical Specifications on ESF Cleanup Systems." Page 25
REFERENCES (DUKE ENERGY CORPORATION RESPONSE)
- 1.
Letter, M. S. Tuckman to USNRC, "Proposed Amendments to the Facility Operating License and Technical Specifications to Allow Insertion of Mixed Oxide (MOX)
Fuel Lead Assemblies and Request for Exemption from Certain Regulations in 10 CFR Part 50," dated February 27, 2003.
- 2.
Letter, Duke Energy Corporation to USNRC, "Proposed Amendment for Partial Scope Implementation of the Alternate Source Term and Proposed Amendment to Technical Specifications (TS) 3.7.10, Control Room Area Ventilation System, TS 3.7.11, Control Room Area Chilled Water System, TS 3.7.13, Fuel Handling Ventilation Exhaust System, and TS 3.9.3, Containment Penetrations," dated December 20, 2001.
- 3.
Letter, C. P. Patel (USNRC) to G. R. Peterson (Duke Energy Corporation), "Catawba Nuclear Station, Units 1 and 2, RE: Issuance of Amendments (TAC Nos. MB3758 and MB3759", dated April 23, 2002.
- 4.
Letter, M. S. Tuckman (Duke Energy Corporation) to USNRC, "Catawba Nuclear Station Units 1 & 2, McGuire Nuclear Station, Units 1 & 2, Mixed Oxide Fuel Lead Assembly License Amendment Request," dated September 23, 2003.
- 5.
Letter, R. E. Martin (USNRC) to M. S. Tuckman (Duke Energy Corporation), "William B. McGuire Nuclear Station Units 1 and 2 and Catawba Nuclear Station, Units 1 and 2, RE: Mixed Oxide Lead Fuel Assemblies (TAC NOS. MB7863, MB7864, MB7865, AND MB7866)," dated July 25, 2003.
- 6.
International Commission on Radiological Protection, 1990 Recommendations of the International Commission on Radiation Protection, ICRP Publication 60, November, 1990.
- 7.
W.D. Travers to the Commissioners; SECY 01-0148, "Processes for Revision of 10 CFR Part 20 Regarding Adoption of ICRP Recommendations on Occupational Dose Limits and Dosimetric Models and Parameters," August 2, 2001. Page 26
- 8. L. Vietti-Cook to W.D. Travers, "Staff Requirements -
Processes for Revision of 10 CFR Part 20 Regarding Adoption of ICRP Recommendations on Occupational Dose Limits and Dosimetric Models and Parameters," April 12, 2002.
- 9. Final Rule on the "Use of Alternative Source Terms at Operating Reactors," (10 CFR Parts 21, 50, and 54; RIN 3150-AG12); FR Vol. 64, No. 246; 71990), December 23, 1999.
- 10. Final Rule on the "Reactor Site Criteria Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants," (10 CFR Parts 21, 50, 52, 54, and 100; RIN 3150-AD93; FR Vol. 61, No. 239; 65157), December 11, 1996.
- 11. Peter S. Tam (USNRC) to William R. McCollum (Duke Energy Corporation), "Issuance of Amendments -
Catawba Nuclear Station Units 1 and 2 (TAC Nos. M98107 and M98108)," April 29, 1997.
This letter announced the issuance of Amendments 159/151 to Facility Operating Licenses NPF-35 and NPF-52 for Catawba Nuclear Station.
An enclosed safety evaluation report presented the findings of the Staff review of a newly discovered design basis accident scenario -
design basis Steam Generator Tube Rupture with failure of control power to the Power Operated Relief Valves of two intact Steam Generators.
The enclosed estimates of upper bounds to the radiation doses for this design basis accident showed that the radiological consequences would be within the acceptance criteria in the license basis at that time (Standard Review Plan 15.6.3 and 6.4.11) if the following conditions were put into effect in place of Technical Specification 3.4.16 (referring to the current Technical Specifications):
Equilibrium reactor coolant dose equivalent 1131 (DEI) specific activity does not exceed 0.46 Ci/gm and Transient RCS DEI specific activity does not exceed 26 pci/gm.
The Staff imposed these limits on operations at Catawba as one of the two License Conditions of Amendment 159/151.
The other License Condition is that an analysis of radiological consequences of the design Page 27
basis Steam Generator Tube Rupture (with the limiting failure(s) for radiological consequences be completed and submitted to the Staff for review before the limits on RCS DEI specific activity in the above License Conditions can be removed.
Currently, this analysis is in preparation.
- 12. USNRC, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants, Regulatory Guide 1.194. Page 28