View source for 2CAN029807, Forwards Response to Telcon RAI Re TS Change Request, Reducing RCS Flow Requirements & Main Steam Isolation Signal Setpoints to Allow Up to 30% SG Tube Plugging.Proprietary Dose Evaluation,Encl.Proprietary Info Withheld

{{Adams
| number = ML20217P425
| issue date = 02/27/1998
| title = Forwards Response to Telcon RAI Re TS Change Request, Reducing RCS Flow Requirements & Main Steam Isolation Signal Setpoints to Allow Up to 30% SG Tube Plugging.Proprietary Dose Evaluation,Encl.Proprietary Info Withheld
| author name = Mims D
| author affiliation = ENTERGY OPERATIONS, INC.
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000368
| license number = 
| contact person = 
| document report number = 2CAN029807, 2CAN29807, NUDOCS 9803100246
| package number = ML19317C961
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE
| page count = 36
| project = 
| stage = Other
}}

=Text=
{{#Wiki_filter:I y                                                                                          Entergy operations,Inc.
1448 SA 333 i                                                                                                RtsseMio, AR 72801
!                                                                                               Tel 501858 5000 l                                                                                                                           =
February 27,1998 2CAN029807 l            U. S. Nuclear Regulatory Commission i           D=maat Control Desk Mail Station OPI-17 Washington, DC 20555

==Subject:==
Arkansas Nuclear One - Unit 2 l                         Docket No. 50-368 l                         License No.' NPF-6 i
Additional Information Corsining the Technical Specification Change Request to Reduce the Reactor Coolant System Flow Requirements and the Main Steam Isolation Signal Setpoint Gentlemen:                                                                                                      I l          - By letters dated September 23,1997 (2CAN099701 and 2CAN099703), Entergy Operations
!            requested changes to the Arkansas Nuclear One (ANO) Unit 2 Technical Specifications.-
l           These changes reduce the reactor coolant system flow requirements and the main steam isolation signal setpoints to allow up to 30% steam generator tube plugging In subsequent conversations with the staff concerning these change requests, additional information j           concerning these amendment requests was determined to be necessary. The anachmaats to i           this letter contains the additional information.
AMechet 3 of this letter contains a proprietary copy of the requested ANO-2 Cycle 12 HZP MSLB Dose Evaluation, Revision 2, dated July 1997. An affidavit from ABB-Combustion Engineering, Inc. is included in Attachment 4 of this letter. The affidavit sets forth the basis l           on which the information may be withheld from public disclosure by the NRC and addresses the considerations listed in 10 CFR 2.790(b)(4). Accordingly, Entergy Operations requests that the topical report contained in this letter to be withheld from public disclosure.                      ,j L           The' attached information does not affect the conclusions of the no significant hazards determination of the September 23,1997 submittals.
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r-U. S. NRC Febmary 27,1998 2CAN029807 Page 2 l
Very tmly yours, l
k&WA Dwight C. Mims l    Director, Nuclear Safety DCM/rde Attachments cc:    Mr. Ellis W. Merschoff Regional Administrator U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Arkansas Nuclear One P.O. Box 310 London, AR 72847 Mr. William D. Reckley NRR Project Manager Region IV/ANO-1 & 2 U. S. Nuclear Regulatory Commission NRR Mail Stop 13-H-3 One White Flint Nonh 11555 Rockville Pike Rockville, MD 20852 I

Attachment I to 2CAN029807 Page 1 During the NRC review of the ANO-2 Technical Specification (TS) change requests to reduce the Reactor Coolant System (RCS) flow and the main steam isolation signal (MSIS) setpoints to allow up to 30% steam generator tube plugging, some additional information was requested. The additional information is contained within this letter with the exception of completing the review of the safety evaluations listed in Attachment 2. The results of the review of the NRC safety evaluations to ensure compliance with the associated conditions will be transmitted to the NRC at a later date under a separate cover letter as discussed under item 7 below.
I
: 1. The NRC requested references to NRC reviews of the HERMITE and ROCS computer                      !
codes.
The requested 1-D HERMITE reference for use in the 4-pump loss of flow event is included in {{letter dated|date=July 13, 1987|text=letter dated July 13,1987}}, to the NRC from Palo Verde Nuclear Generating Station. The staff accepted this request in the evaluation dated August 25,1987.
The requested ROCS and HERMITE reference for use in the Main Steam Line Break (MSLB)                 I analysis is included in the safety evaluation for the Waterford 3 reload analysis for Cycle 2 dated January 16, 1987. The Waterford 3 submittal, dated October 1,1986, references the Calvert Cliffs Unit I submittal dated September 1,1983, for its Cycle 7 MSLB analysis.
Approved Topical Reports:
* CENPD-188-A, "HERMITE A Multi-Dimensional Space-Time Kinetics Code for PWR Transients" e  CENPD-266-P-A, "The ROCS and DIT Computer Codes for Nuclear Design"
: 2. The NRC requested an explanation of the use of the HRISE computer code.
The ANO-2 Safety Analysis Report (SAR) currently indicates the use of the MacBeth Correlation by reference 19 to Chapter 15 as follows:           MacBeth, R. V., " Burnout Analysis Part 4:
I Application of a Local Conditions Hypothesis to World Data for Uniformly Heated Round Tube and Rectangular Channels," AEEW-R-267, August 1%3. ANO-2 has used the MacBeth Correlation since the original licensing analysis. HRISE is an automated version of the MacBeth Correlation. Additional documentation can be found in the Calvert Cliffs submittal dated March      -)
28,1983, and the associated SER dated July 15,1983.                                                  I
: 3. The NRC mquested an explanation of the assumed HPSI and LPSI flows for the small                  >
bnak LOCA analyses.
The Emergency Core Cooling System (ECCS) performance analysis for the small break loss of            ;
coolant accident (LOCA) assumes a cold leg break and the limiting single failure of one of the Engineered Safety Feature (ESF) diesel generators. This single failure minimizes the availability of flow from the High Pressure Safety Injection (HPSI) and the Low Pressure Safety Injection

Attachment I ts 2CAN029807 Page 2 (LPSI) systems. SAR Figure 6.3-2 contains a drawing of these systems and the associated               i injection header motor operated valves (MOVs).
The HPSI system is Ngaad with two separate and independent trains. Each train consists of one pump and four injection lines, one to each of the four RCS cold legs. Each of the four injection lines contain a normally closed injection MOV that is powered from the same ESF diesel generator as the pump. A loss of an ESF diesel generator results in a complete loss of the associated train of HPSI t-- of the loss of power to the pump and its injection MOVs. The remaining train is assumed to operate as designed with the pump starting and its injection MOVs opening The flow from one of the four injection lines (25% of the total flow) in the operating HPSI train is assumed to be lost out the cold leg break. The remaining 75% of the HPSI flow enters through the other three injection lines and is assumed to be available for core cooling.
The LPSI system configuration is different from that of the HPSI system. The LPSI system consists of two pumps discharging into a common header that feeds four injection lines. Each of the injection lines direct flow to each of the four cold legs. There is a normally closed MOV in each of the injection lines. Two of these MOVs are powered by one ESF diesel generator and the remaining two MOVs are powered by the other diesel generator. The loss of one of the ESF diesel generator results in the loss of the associated LPSI pump and the loss of two MOVs which       {
are assumed to be closed. The remaining pump is assumed to operate as designed, delivering flow through the remaining two open injection MOVs into their associated cold legs. One of these two operating injection lines is assumed to be connected to the ruptured cold leg. The flow from the ruptured cold leg (50% of the total flow) is assumed to be lost out the break. The remaining 50%
of the LPSI flow enters through the other injection line and is assumed to be available for core cooling.
: 4. The NRC requested a description of the resolution of the identified error in the energy redistribution factor used in the large break LOCA analysis.
As identified in the ABB-Combustion Engineering (ABB-CE) 10 CFR 21 repon to the NRC (I. C. Rickard of ABB-CE to the NRC, " Report pursuant to 10 CFR 21 Regarding Error in the .
Energy Redistribution Factor Used in LOCA Analysis," August 14,1997), the 1975 generated curve of energy redistribution factors (ERF) as a function of pin / box peaking factor neglected to include the effects of voiding that occurs during the LOCA event. Large break LOCA analyses employed this curve in conjunction with a 0.8% allowance for uncertainty. This curve, including the associated uncertainty, has now been regenerated using the MCNP Monte Carlo code.
The current limiting value of the minimum pin / box factor applicable to the LOCA analyses for ANO-2 is 1.03. The ERF used in the current analyses for a pin / box factor of 1.03 is 0.960. With consideration of the voiding effects, the new ERF for a pin /bex factor of 1.03 is 0.969. However, this pin / box factor is a conservative minimum applicable to a range of cycle designs. The actual ANO-2 Cycle 13 pin / box factor is 1.0632.          The ERF for this pin /bex factor, including considerations of the voiding effects, is 0.956. The ERF used in the curren large break LOCA analyses is larger than this ERF, and is thus, conservative. As such, there is nc need to update the LOCA analysis.
However, due to the timing of this issue, additional compensatory actions have been taken. A conservatively estimated increase of 40'F in the peak clad temperature (PCT) was originally

l I
Attachment I ts 2CAN029807 Page 3 considered. To compawe for the error in the ERF, the linear heat rate for Cycle 13 has been reduced by 0.2 kw/ft, from 13.5 kw/ft to 13.3 kw/ft. The revised Core Operating Limits Report          1 (COLR) was submitted to the NRC on August 20,1997 (2CAN089703). This reduction in the linear heat rate more than compensates for the maximum 40 F change in PCT, and reduces the calculated PCT to 2158'F, without consideration of the cycle specific pin / box factor.
The reduced limit on linear heat rate will apply to the remainder of Cycle 13. The reduction in the linear heat rate limit to 13.3 kw/ft is not expected to create any operational difficulties for the remainder ofCycle 13 or for Cycle 14.
  ~ Because of the known conservctisms in the LOCA analyses, the large break LOCA analyses will not be redone at this time. The large break LOCA analyses will be redone to support the replacement of the steam generators for Cycle 15. This reanalysis is expected to be completed in the Fall of 1999 and will include updated ERFs to correct tne identified error to restore margin in the allowable pin to box factor.
: 5. The NRC requested a basis for transition from HSTA code to RELAP for calculation of feedwater flow for the NSSS impact evaluation of a MSLB.
1 The original analyses of the impact of a main steam line break (MSLB) on the primary system employed the HSTA code to calculate the feedwater flows to the ruptured and intact steam                !
generators following the break. The HSTA code has been replaced by the more sophisticz.ted              l RELAP code in the new analyses. RELAP has been accepted by the NRC for use in numerous safety analysis applications, including the modeling of the feedwater flow follow ~ng a MSLB for containment impact evaluation at ANO Unit 2.
For the revised MSLB analyses, for both primary system and containment impacts, the feedwater flow rates to the steam generators were calculated using RELAP5 MOD 3.1 for a series of active
!  component single failures. The new RELAP calculations have produced more conservative l  results than those from the HSTA code. In general, RELAP predicts higher feedwater flows throughout the transients with higher integrated flows at all critical time points, including the peak return to power, peak reactivity and minimum DNBR. The prediction of higher integrated flows assures that the evaluation of the MSLB impact on the primary system will be conservative.
: 6. The NRC requested a comparison of analysis input parameters in the RCS flow reduction technical specification amendment request and Amendment 13 to the ANO-2 SAR.
The input assumptions identified in Attachment I to the technical specification amendment request have been compared to those listed in Amendment 13 of the ANO-2 SAR. The following discussion addresses any differences between the assumed values. The lists of significant parameters are not always the same for both documents, primarily as a result of additional detail provided in the documentation of the newer analyses. Only those parameters listed in both              i documents, with different assumed values, will be addressed.
The amendment request groups all significant input assumptions into tables. The comparison of parameters was based on the tables and the discussion of the differences will be presented in a table by table format. The table number designations given in this attachment correspond directly J
ts 2CAN029807 Page 4 with those of the Attachment I to the amendment request. Although some input assumptions are incorporated into the text, most of these values are repeated in the tables. The assumptions mentioned in the text that are not included in the tables are generally related to physics parameters that will be addressed generically.
The two most significant sources of differences between the SAR and amendment request assumptions are the reduction in RCS flow itself and the use of current cycle physics data. The plugging of steam generator tubes and the resulting reduction of RCS flow also impact the normal ranges of other primary and secondary system parameters. In the analyses presented in the amendment request, the effects of tube plugging was considered on related input parameters. The values of these parameters were set to produce the most conservative results. The bases for the differences in the values of RCS flow and core flow are explained in the amendmet request; these differenca will not be included in this letter.
The analyses used physics parameters from recent operating cycles, primarily Cycle 13, the current operating cycle. The values used for these parameters are generally expected to be bounding for the remaining cycle (Cycle 14) prior to the replacement of the steam generators.
Consistent with the ANO established reload methodology, these parameter assumptions are reviewed for the subsequent cycle, and if wa==y, operating limits are adjusted to assure that the analysis results remain bounding.
: 7. The NRC requested confirmation that the conditions specified in the NRC safety evaluations for the topicals associated with the analyses that were performed in support of the RCS flow and MSIS setpoint reduction efforts were satisfied.
The analyses supporting the technical specification changes for the RCS flow and MSIS setpoint reduction efforts were prepared by ABB-Combustion Engineering (ABB-CE). It is ABB-CE's policy to be in compliance with the requirements contained in the approved methodologies and with the requirements contained in NRC prepared Safety Evaluation Reports (SERs). In response to this question from the NRC, Entergy requested ABB-CE to review the safety evaluations for the topicals associated with the analyses that was performed in support of the MSIS and reduced RCS flow TS changes and to identify the conditions specified in those documents. Following identification of the conditions, AAB-CE will verify the conditions were met for the analyses performed in support of the ANO-2 TS changes.
This review is currently underway and is expected to complete on or about March 4,1998.
Entergy will transinit the requested information to the NRC following internal review. The SERs which are being reviewed are listed in Attachment 2 of this letter.
This request to document compliance with the conditions of the associated SERs was not a previously established expectation from the NRC for a licensing submittal. It is our understanding      l that this request to document compliance with the conditions of the SERs is a result of the             ;
recently developed generic NRC concern in this area. Because the majority of the industries' safety analyses are performed by the nuclear steam supply (NSS) vendors (in ANO-2's case by ABB-CE) and because most of the analytical tools are being applied generically to the licensees serviced by these NSS vendors, Entergy plans to bring this issue before the CE owners group for consideration. Likewise, we are communicating with the Nuclear Energy Institute (NEI) because a

Attachment I to 2CAN029807 Page 5 of the generic implication across vendor lines. To fully address this issue it would be beneficial to have fmther dialogue with the NRC on this topic.
: 8. The NRC requested a comparison of the HSTA and the RELAP5 MOD 3.1 codes for the                    l main steam line break over cooling events.
The original main steam line break reactor over-cooling analyses for ANO-2 used the HSTA code to calculate feedwater flow to the steam generators after the break. The revised analyses for Cycle 13, which includes the reduced RCS flow, used a RELAPS (specifically RELAP5 MOD 3.1) model of the feedwater system to calculate feedwater flow to the steam generators after the break. The RELAPS model is more versatile and sophisticated than the HSTA code, and provided greater flexibility in the evaluation of postulated single failures for the new analyses.
To demonstrate that the RELAP5 model produces conservative results compared to the HSTA medel, the integrated feedwater flows to the ruptured steam generator, calculated by both models, are shown in the attached figures Figure 1 presents the integrated flows for the loss of AC case and Figure 2 presents the data for the AC available case. A comparison ofintegrated feedwater flows directly demonstrates the relative conservatism of these two models. Feedwater flow to the ruptured generator efter the break increases the total energy removed from the primary system, which in turn increases the RCS cooldown and the associated reactivity insertion. Therefore, higher feedwater flow produces more conservative results.
The data presented in Figure 1, for the loss of AC case, is the best demonstration of the              l conservatism in the RELAPS results. The HSTA and RELAPS model results are both for the                 !
single failure of the main feedwater isolation valve to close. The integrated flow predicted by        )
RELAPS is higher than the HSTA results at all times during the transient.
The data presented in Figure 2, for the AC available case, represent different assumed single failures. The HSTA results are for the single failure of a condensate pump to trip. The RELAPS results are for the single failure of a main feedwater pump to trip. The condensate pump and main feedwater pump single failures were considered for both models. HSTA determined the condensate pump failure to be more limiting, and RELAP5 determined the main feedwater pump failure to be more limiting. Figure 2 compares the limiting results for both models for these assumed failures. Again, the RELAPS integrated flow is higher at all points during the transient.
: 9. The NRC requested a review of topical report CENPD-384-P to determine if it is applicable to ANO-2.
The conclusions of ABB-CE Topical Report CENPD-384-P, " Report on the Continued Applicability of 60 MWD /kgU for ABB-Combustion Engineering PWR Fuel," are directly applicable to ANO-2. The topical report considered the behavior of high burnup fuel under reactivity insertion transients. The topical report assessed generic ABB-CE PWR fuel and concluded that no further plant specific evaluations were required. ANO-2 has always used and continues to use fuel supplied by ABB-CE.

A*=A-d I to 2CAN029807 Page 6 10.The NRC requested ANO to provide a discussion explaining why the maximum hypothetical accident is the bounding event for control room dose.
Consistent with the licensing basis for ANO-2, the control room emergency ventilation system (CREVS) effectiveness is evaluated on the basis of the maximum hypothetical accident (MHA) control room dose. The MHA control room dose is demonstrated to be less than the limits of GDC 19 and will bouw$ t'.se doses from a main steam line break (MSLB).
The conservatism in this conclusion can best be demonstrated by a comparison of the calculated release rates of dose equivalent Iodine-131 for equivalent time periods of the MHA and MSLB events. A comparison ofiodine releases is pertinent because the thyroid dose is significantly more limiting than the whole body or skin doses; this will be tme for both the MSLB and the MHA events.
The total number of curies of dose equivalent iodine released at 2 hours and 8 hours following both events are presented in the following table. The release rates for the MHA are from the current analysis for ANO-2 and include only the releases due to containment leakage. The release rates for the MSLB are from the dose calculation referenced in the technical speci6 cation change request (2CAN099703), and include the effHis ofiodine spiking. The values presented with
,    respect to MSLB are the higher release f. . W -r the pre-existing or event generated iodine spike cases.
Release ofDose                Ma 4 . s rn etical        Main Steam Line Break Equivalent Iodine-131                    :M c                          Accident at 2 hours                      289 Curies                     26 Curies at 8 hours                       576 Curies                    156 Curies This comparison demonstrates that the MHA release is more than ten times as large as the MSLB release in the first two hours, and almost four times the MSLB release at the end of eight hours, at which time the MSLB release is assumed to end.
Apart from the difference in iodine release rates, other aspects of the MHA control room dose calculation would be essentially the same for a calculation of the MSLB dose. Control room isolation and actuation of the CREVS would be the same for both events. For both events, control room isolation and CREVS actuation would be initiated by a high radiation signal from either one of the radiation monitors in the air supply ducts of the Unit I and Unit 2 normal HVAC systems. (Details on the Unit I duct monitor are provided below.) The location of these monitors '.n the normal air supply ducts and their sensitivity assures that they will very quickly be exposed to and respond to activity released outside containment from either event. The rapid response of the monitors is assumed for either event even thout,1. the activity release phenomena for the two events are different.
For the MHA, activity is released from the RCS into the containment. From there, the activity is assumed to be released to the environment around the containment at a rate determined by the contaimaent leakage limits. For the MSLB, activity is released from the main steam line at a point just outside containment. From tinis release point, all activity is assumed to escape directly to the

Attachment I to
  - 2CAN029807 Page 7 atmosphere. Although, as described above, the activity released outside of containment during the MHA is si sfwiely larger than for the MSLB, the initial build up of activity outside containment will be quicker for the MSLB than for the MHA due to the extra transport time through the containment and out the leakage paths, Consequently, activity released from the MSLB will reach the control room normal air supply intakes and the duct monitors faster than activity released from the MHA.
The rate at which activity concentration at the control room air intake develops following a MSLB is an extremely complex function of meteorological conditions, break location and configuration, break release mass and energy, and the physical configuration of the plant.
However, for the MSLB as well as the MHA, the build up of activity levels at the control room air intake is expected to be gradual, reaching peak levels in minutes rather than seconds. The initial blowdown of the ruptured steam generator lasts for more than a minute and a half. The velocity and temperature of the initial release will be high resulting in a turbulent, rapidly expanding (vertically as well as horizontally) cloud that will disburse the activity rather than concentrate h at the intake.
The location of the duct monitors assures that the raonitors will sense the increased activity before it reaches the control room envelope. Once the released activity reaches the monitor, the high sensitivity of the monitor will quickly produce an actuation signal at an activity concentration that is very low (i.e. twice background). The isolation of the control room will be complete within 10 seconds of this actuation signal. Although the activity concentration being drawn into the normal air intake will continue to increase during the short time required for the monitor to respond and the control room to isolate, the total activity reaching the control room prior to isolation is expected to be small due to the gradual concentration incr ease at the air intake.
Consequently, in the same manner as the MHA analysis, the analysis of the MSLB control room dose would assume the immediate isolation of the control room and the actuation of the CREVS.
The small dose contribution from activity drawn into the control room prior to isolation is considered negligible compared to the doses received during the remainder of the transients.
With the morol room isolated and the CREVS in operation, the only assumptions which could potentially differ between the MHA and MSLB control room dose analyses are the x/Q values.
The remaining assumptions (such as breathing rates, dose conversion factors, etc.) and the control room and CREVS operating parameters (such as control room volume, air in leakage, filter                 -
efficiencies, etc.) used in the MSLB control room dose analysis, would be identical to those for the calculation of the MHA control room dose (over the eight hour duration of the MSLB).
On the basis of engineeringjudgment, the assumed x/Q values used for the hC4A analysis are also assumed to be i yr        aative for the MSLB, even though the releases would occur differently.
l The MSLB produces a singic release point outside the containment, while containment leakage from the MHA is assumed to occur all around the building. ~ Despite these differences, the relative locations of the main steam lines exiting containment (east northeast side of containment) and the anticipated pathways to the control room emergency air intakes (south east of containment) suggest the diffusion pathway for activity from the single MSLB release point would not differ significantly from the average pathway for MHA leakage points around the containment.
Although some differences may exist in the MHA and MSLB x/Q values, the effect of those
____m

Attachment I to 2CAN029807 Page 8 1
differences on the control room dose is expected to be much smaller than the dose differences due to the disparity in the total curies released for the MHA and MSLB.
With the location and sensitivity of the control room radiation monitors supporting the assumption that control room isolation effectively occurs immediately for both the MHA and MSLB, and with all other dose analysis assumptions and inputs the same, the substantially larger release term for the MHA assures that the calculated control room dose for the MHA will bound that of the MSLB. The MHA control room dose has additional factors that make it even more limiting; specifically, the releases from ECCS component leakage and passive component failures are signi6 cant contributions (about one third of the total MHA dose) that would not be applicable to the MSLB dose. Also, the MHA release is not assumed to be terminated at 8 hours, but is extended for a total of 30 days.
: 11.      The NRC requested ANO to provide a discussion of the new ANO-1 control room intake duct monitor.
The ANO-1 and ANO-2 control rooms are located adjacent to each other within a common control room envelope. The current ANO. I control room radiation monitoring system consists of an area radiation detector and associated display unit which monitors the control room and provides isolation and annunciation signals on the detection of high radiation. The existing ANO-2 control room radiation monitork.c system consists of a radiation detector installed in the ANO-2 control room supply air duct. The radiation detector located in the ANO-2 normal air supply duct also has an associated display unit which monitors the control room supply air and provides isolation and annunciation signals on the detection of high radiation. If either of these radiation monitors reach their associated isolation setpoint, the control room isolation dampers will be actuated closed and the associated train of CREVS will be started. This condition of control room isolation and filtration is called emergency recirculation.
A plant modification is being installed that will modify the existing control room radiation monitoring system. This modification adds an additional radiation detector in the ANO-1 normal air supply duct which will supplement the existing area monitor. The system will be configured so that if either the area monitor or the new ANO-1 duct monitor senses radiation above their respective setpoints, the control room will be automatically placed on emergency recirculation.
This modification effectively allows both of these radiation monitors to operate in parallel with the capability of placing the control room on emergency recirculation. The new ANO-1 duct monitor consists of a scintillation type detector, and an associated preamplifier and rate meter with a sensitivity that is the same as the ANO-2 control room radiation monitor. The new duct monitor also will have the same s 2 times background isolation setpoint requirements as that of the existing ANO-2 duct monitor.
All of the components installed by this modification were procured as nuclear qualified and mounted seismically with all seismic II/I issues addressed. The new duct monitor is also powered from a diesel backed safety grade (class IE) power supply. These components will be the same type of equipment that is installed in the ANO-2 control room normal air supply duct. The existing ANO-1 control room area radiation monitor is not designed as a safety grade component and is powered from a non-class IE power supply. The new ANO-1 duct monitor improves on the current design by installing safety grade equipment that is powered from a diesel backed IE

Attachment I to 2CAN029807 Page 9 power supply. However, both the existing area monitor and the new duct monitor will actuate the safety grade isolation relay via cables that are not class IE consistent with the current design. The new duct monitor simply taps into the existing non-class IE circuit from the control room area monitor circuit. Electrical isolation is provided between the non-safety related detection circuit and the safety related components.
The new ANO-1 duct monitor is expected to be placed in service prior to the issuance of the            !
MSIS setpoint reduction TS that was requested by letter 2CAN099703 dated September 23, 1997. ANO-2 TS surveillance requirement 4.7.6.1.2.d requires two trains of radiation detection for control room isolation. ANO-2 currently requires the operability of the ANO-1 area radiation      q monitor for one of the trains and the ANO-2 duct monitor for the other train. Once this modification is complete, ANO-2 will require the operability of the new ANO-1 duct monitor and
..ill no longer credit the existing ANO 1 area monitor for control room dose assessment and TS compliance.
By letters OCAN049501 dated April 4,1995, and letter OCAN129603 dated December 19,1996, Entergy Operations requested changes to the ANO-1 and ANO-2 TS for the common control room radiation detection, emergency filtration, and emergency air conditioning equipment. These TS change requests essentially made the TS operability and surveillance requirements for this equipment the same for both units. Entergy Operations will submit an additional modification to these submittals by August 1,1998 to address the new ANO-1 normal supply duct radiation monitor. Entergy Operations will also administratively app'y the same surveillance and operability requirements that currently exist for the ANO-2 duct monitor, to the ANO-1 duct monitor until the above referenced TS change requests are approved. In addition, Entergy Operations also            !
requests the NRC to increase the priority on the staffs review of above referenced TS changes for     ,
the common control room radiation detection, emergency filtration, and emergency air conditioning equipment. These changes were proposed to eliminate the differing requirements           i located in both units TS for the common control room equipment.
: 12.      The NRC requested ANO to prwvide a discussion of the Core Protection Calculator (CPC) rdter verification analyses performed for ANO-2.                                        i I
The CPC filter verification analyses were performed to ensure the acceptable performance of certain CPC digital processing filters at ANO-2. These filters are included to ensure that the CPC protective calculations are provided with conservative inputs taking into account the actual rates of change ofNSSS parameters that occur during certain DBEs.
The objective of this analyses was to conservatively tune the CPC transient compensation iZtres.
These filters are:
Increasing Power Filters Increasing Temperature Filters Dec. easing Temperature Filters I

i Attachment I ts                                                                                     I 2CAN029807 Page 10 The para.neters of primary *anportance to these filters are not generally subject to change due to reload core ^=7 However, when a significant perturbation to plant configuration is planned, such as the current tube plugging /RCS flow reduction effort, the CPC filter verification is checked.
A discussion ofevents applicable to each of the above filter categories follows.
Increasing Power Events Of the possible increasing power events, Boron Dilution, Excess Loads, Single and Bank CEA Withdrawals, the Bank CEA Withdrawal event is examined because it produces the fastest rate of power increase. The rate of increase of CPC compensated neutron flux power (PHICAL) is compared to the actual rates ofincrease of both core power and core heat flux to ensure that the inputs to the CPC protective calculations are conservative.
Increasing Temperature Futers Several potential decreasing heat removal anticipated operational occurrences (AOOs) are past of the CPC design basis events. The Lou of Feedwater event and the Loss of Load event each decrease the rate of heat removal. The, Loss of Load event results in the more adverse heatup -
scenario.
The CPC filters provide two RCS temperature indications which are used in the protective calculations. These are TCMIN and TCMAX.
TCMAX is used as the core inlet temperature for the CPC DNBR calculation.                   To be conservative, the value of TCMAX must always be larger than the ' actual' core inlet temperature.
TCMIN is used to adjust the raw neutron power signal for downcomer temperature shadowing, a correction term known as Tr, where Tr = 1.0 + C*( Tref- TCMIN) and the parameter 'C' has different values for TCMIN greater than or less than Tref.
To be conservative, the value of TCMIN should be less than the actual RCS temperature existing in the downcomer region of the reactor vessel at any given time in the transient.
The CPC lead and lag temperature filters adjust the input RTD to obtain the desired conswatively low TCMIN and conservatively high TCMAX. The Loss of Load is selected as the CPC DBE to examine these filters for transients resulting in increasing in RCS temperatures.
Decreasing Temperature Filters A spectrum of possible excess loads could be imposed upon the NSSS by the secondary system.
An increase in feedwater flow or decrease in feedwater temperature has the potential to increase the heat demand on the primary system. A single ADV or turbine bypass valve have the capability of causing an increase of ~10% of rated steam flow. A change in the position of the turbine        l l
l

Attachment I to 2CAN029807 Page 11 admission valve has the potential to result in larger variations in the steam demand from the initial conditions.
A spectrum of potential turbine admission valve driven increases in steam flow was analyzed to ensure that the filters are conservative in the decreasing temperature direction over the spectrum of possible power to load imbalances.
Method of Analysis The transients discussed above are simulated with the CENTS simulation code. The time dependent plant parameters calculated using the CENTS code are then input into the CPC simulation code (referred to as 'CPC FORTRAN').
The output from the CPC FORTRAN code is then compared to CENTS ' reality' to judge the conservatism of the CPC filtering process.                                                              I Criteria for Analysis The CPCs are designed to result in a reactor trip early enough to prevent SAFDL violation for the above set of AOOs. This analysis must demonstrate conservatism resulting from the CPC filtering l process for several important process parameters. This in combination with the appropriate              I l setting of other CPC parameters (i.e. core specific Fxy values, Rod Shadowing values, etc.)              I ensure timely CPC action.
In the specific case of the CEA Withdrawal event in the current analytical effort, the response of the CPCs to the transient simulation of core power and heat flux from the CENTS case are l compared. The CPC power, PHICAL, is compared against CENTS power (for LHGR acceptability) and CENTS heat flux (for DNBR acceptability)in the Figure 3 of this attachment.
It is seen that CPC power is essentially equal to or larger than CENTS power and heat flux as indicated on Figure 3. This illustrates acceptable functioning of the CPC filters.
For a CEA Withdrawal event which starts at a condition of minimum thermal margin, as preserved by COLSS, and proceeds through the transient essentially 'using up' all of the thermal margin, a typical transient response of DNBR and thermal margin versus time is provided on the two Figures 4 and 5 of this attachment. Due to the CPC filters, the calculation of DNBR in CPC will be based on a conservatively high heat flux and, therefore, a CPC trip will be generated before minimum DNBR decreases to the SAFDL.

1 A'*Me I to 2CAN029807 Page 12 Table 1 System Paranseters And Initial Conditions for The Large Break LOCA ECCS l      Performance Evaluation with Increased Tube Plugging And Reduced RCS Flow Rate
{
!                                                                                                      l
;                                                        Amendment            SAR Request         Assumed                     j Ouantity                      Assumed Value         Value          Ilni11 Gap conductance at the PLHGR(U                         2136              1572       BTU /hrft2 .p l
Fuel centerline temperature at the PLHGR(4             3204              3359             'F Fuel average temperature at the PLHGR(4                 1984             2129             'F      i l
Hot rod gas pressure (U                                2647              1114            psia Hot le3 temperature                                    622.7            616.8             *F (D
These quantities correspond to the rod average bur.:up of the hot rod that yields the highest peak cladding temperature.
The differences in the first four parameters are the result of assuming a rod average burnup of 40,000 MWD /MTU in the amendment request analyses as opposed to a burnup of 1000 MWD /MTU which was the basis of these parameters in the SAR analyses. The assumption of the higher burnup produces more conservative analysis results.
The increase in hot leg temperature is the direct result of decreasing the RCS flow while maintaining the core inlet temperature at its maximum value.                                      j Table 7 System Parameters And Initial Conditions For The Small Break LOCA ECCS                      )
Performance Evaluation                                       l The presentation of the small break LOCA analyses was included in the Amendment Request for completeness These analyses have already been approved as a part of Technical Specification Amendment 179.

A++d-aa* 1 to 2CAN029807 Page 13                                                                                   i Table 12                                             .
Assumptions For The Uncontrolled CEA Withdrawal From A Suberitical Condition Amandment Request               SAR Assumed Value              A-mad 1
Parameter                 Haita           .Cang_1         Case 2           Value Initial Core Power                   (MWt)         8%.0 x 10*      896.0 x 10*      253.0 x 10'''
l Core Inlet Temperature                 ('F)             552             552              545        i Steam Generator Pressure              (psia)            1055           1055             1004 CEA Maximum Reactivity            (10dAp/sec)            2.5            2.0              4.0 Addition Rate
  - SAR Table 15.1.1-1 incorrectly gives a value of 2.53 x 10-'' for the initial core power. The typographical error is corrected in the table above. The smaller initial power used in the SAR analysis is due to the assumption of a source term based on a 163 day outage, while the             !
amendment request analysis source term is based on a more representative 100 day outage. A conservatively small value for the initial power is chosen, consistent with the 100 day outage     ,
assumption, to ensure conservative analysis results.                                              j The core inlet temperature used in the SAR analysis is a nominal value. Although the analysis results are not highly sensitive to this parameter, a more conservative value was used in the amendment request analyses. The higher temperature is conservative with respect to DNBR considerations. The increased steam generator pressure is calculated during the initialization of the CENTS analysis and is the direct result of the increased core inlet temperature assumption.
The maximum reactivity addition rate retlects updated physics parameters considered bounding for Cycle 13 and future cycles. Credit is also taken for procedural controls on the maintenance of subcritical margin which limit the possible combination of rods that can result in criticality during an uncontrolled withdrawal. Note that the units listed in the SAR for the CEA reactivity addition rate are incorrectly given as 10-2 Ap/sec; the units in the table above are correct.

Attachment I to 2CAN029807 Page 14 Table 15 Assumptions For The Uncontrolled CEA Withdrawal From Hot Zero Power Amendment       16R Reauest     Assumed Parameter                           Qni.tg         Assumed Value     Value Initial Core Power                                 (MWt)              0.002815        29.0 Steam Generator Pressure                            (psia)               1055         1050 CEA Worth on Trip                                  (% Ap)                 -2          -5.5 CEA Maximum Reactivity Addition Rate           (10" Ap/sec)               1.8          1.5 The initial core power assumed in the SAR analyses was a nominal 1% value. The 4
amendment request analyses use a lower value (10             %) which produces slightly more I conservative results by extending the transient to produce a larger power spike.
The initial steam generator pressure is calculated during the initialization of the CENTS analysis and is essentially the same as the SAR value.
The CEA worth on trip in the amendment request analyses is a conservatively bounding value based on updated physics data.
l The maximum reactivity addition rate reflects a bounding updated physics parameter           j assumption.
Table 17                                         i Awumptions For The Loss Of Coolant Flow Analysis Assuming 30%
Steam Generator Tube Plugging Amendment Request Assumed     SAR Assumed Parameter                      Units             Value             Value Moderator Temperature Coefficient         (10d Ap/ F)             0.0              +0.5 Scram Worth                                 (% Ap)              -5.0              -8.0 The SAR assumed value for the moderator temperature coefficient was conservative given the technical specification limitations on the MTC at full power. The assumed value in the amendment request analysis is consistent with the technical specifications and is still conservative.
The CEA worth on trip in the amendment request analyses is a conservatively bounding value based on updated physics data.

Attachment I to '
2CAN029807 Page 15 Table 19 Assumptions For The 14ss Of External Lond/ Loss Of Condenser Vacuum Amendment                 SAR'          ;
Request            Assumed            l Parameter                     Vita         Am-A Value               Value           i i
Steam Generator Pressure                    (psia)              795                _810           !
Moderator Tc:nperature Coefficient      (10" Ap/'F)              0                  +0.5 CEA Worth or. Trip                         (% Ap)              -5.0                 -5.4 The initial steam ge:yxator pres;sMe is calculated during the initializatim of the CENTS analysis and reflects the impacts of Rbe p!ug$ng.
The SAR assumed value for the moderator temperature coetlicient wu conservative given the technical specification limitations on the MTC at full power,          The assumed value in the amendment request analysis is consistent with the technical specifications and is still conservative.
The CEA worth on trip in the amendment request analyses is a conservatively bounding value based on updated physics data.
1 Table 21                                                l Assumptions For The Cycle 13 Loss Of Normal Feedwater Flow                          l l
Amendment                             j Request Assumed          SAR Assumed     j brameter                      Vits                Value                  Value Reactor Coolant System Pressure             (psia)               2000                   2250 Steam Generator Pressure                    (psia)                922                    923 Moderator Temperature Coefficient       (10d Ap/*F)              -3.5                   +0.5 l
l 1,4                             l Doppler Multiplier                            .                                         0.85 I
CEA Worth on Trip                         (% Ap)                 -5.0                     5.4 l
l

Attaa-t I to 2CAN029807 Pag) 16 The initial RCS pressure assumed in the SAR analysis was a nominal veJue. The value assumed in the amendment request analysis is consistent with the allowable range of operating pressures and chosen to produce conservative results. More conservative results are obtained if the reactor trip occurs on low steam generator level rather than on high pressurizer pressure. The lower initial RCS pressure delays the high pressure trip to extend the transient.
The initial steam generator pressure was optimized to produce conservative analysis results and is essentially the same as the SAR initial value.
End of life moderator temperature and doppler coefficients and the' corresponding doppler multiplier are used in the amendment request analysis to produce more conservative results.
Sensitivity analyses have shown that the more negative coefficients will increase.1he post trip core power response and produce slighti/ lower steam generator inventories.
The CEA worth on trip in the amendir.ent request analyses is a conservatively bounding value based on updated physics data.
Table 23 Assumptions for the Steam Line Break Analysis From Hot Full Power and Hot Zero Power Amendment Request           SAR Assumed Assumed Value               Values Hot Full   Hot Zero Hot Full       Hot Zero Parameter _              Units          Power     Power       Power       Power i
Initial Core Inlet Temperature              F            --       552          -
547.6 Initial RCS Pressure                     psia          2300      2300        2250        2250   ;
CEA Worth on Trip                      % Ap         -7.5144        -          -8.6         -
Initial Steam Generator Pressure         psia           922      1058         932         1025 Doppler Coefficient Multiplier            -            1.22      1.22         1.15        0.85  l MTC Multiplier                            -             1.0       1.0          1.1         1.1 MTC                                           4         -3.4      -3.4         -3.5        -3.5 (10 api *F)
The core inlet temperature used in the SAR hot zero power analysis is essentially a nominal value. Although the analysis results are not highly sensitive to this parameter, a more conservative value was used in the amendment request analyses.

A"=^M 1 to 2CAN029807 Page 17 The initial RCS pressure assumed in the SAR analysis was a nominal value. The value assumed in the =M             request analysis is consistent with the allowable range of operating pressures and chosen to produce conservative results. The higher initial pressure delays the start of safetyinjection.
l l    The CEA worth on trip in the amendment request analyses is a conservatively bounding value based on updated physics data.
l l
The steam generator pressure is calculated during the initialization of the CENTS analysis.
For the zero power analysis, the increased pressure is the direct result of the increased core inlet temperature assumption. For the full povr analysis, the initial steam generator pressure was optimized (assuming 0% plugged tubes) to produce conservative analysis results and is essentially the same as the SAR initial value.
;    End oflife doppler coeEcients and a very conservative doppler multiplier are used in the amendment request analyses to produce more conservative results. The doppler coefficient
!     multiplier for the hot full power case is incorrectly listed in the SAR as 1.5.                 This typographical error, introduced in an update from the FSAR, is corrected in the table above.
The original SAR analyses used a moderator temperature coefficient multiplier of 1,1 to l     account for calculational uncertainties. This was applied to the reacti ity versus moderator density function developed for the original SAR analysis. Subsequent analyses, including the amendment request analysis, no longer use this conservatism. Use of the end oflife design limit for MTC still maintains conservative analysis results.
The change in the moderator temperature coefficient from -3.5 x 10" AprF to -3.4 x 10" l     AprF reflects a change in the technical specification limit from Cycle I to the current cycle        i limit.
Table 28 Assumptions For The CEA Ejection Accident Analysis Amendment Reauest            SAR Assumed Parameter                 Units             Assumed Value                 Values
* i                                                          HZE           HEE         HZE         HEE 1
Initial Core Power                  (MWt)              29             -            1           -
Total Delayed Neutron                   -
0.004.'4414 0.0043414      0.00482     0.00482 Fraction (p)                                                                                           )
i d                                                      +0.5 MTC                               (10 AprF)            -             0.0          -
CEA Worth on Trip                    % Ap              -2            -5          -2.4         -5.4
* The SAR values come from SAR Table 15.1.20-8 and the associated text on reload cycle methodology in Section 15.1.20.2.1.2.
l

Anache 1 to 2CAN029807 Page 18 l
The initial core power assumed in the SAR analyses was a nominal 1 MWt value. The l     amendmem request analyses use a higher value which produces slightly more conservative reults.
The analyses for the amendment request used a combination of beginning oflife and end of life (EOL) physics parameters. The difference in the EOL delayed neutron fraction in the arpandmaat request analyses reflects the use of updated physics parameters.
The SAR assumed value for the moderator temperature coefBeient was conservative given the technical speci6 cation limitations on MTC at full power. The assumed value in the amendment request analysis is consistent with the technical specifications and is still conservative.
The CEA worth on trip in the amendment request analyses are conservatively bounding values based on updated physics data.
Table 29 Axial Power Distribution Used For The CEA Ejection Accident Analyses l
Amendment Request Fractional Distance from the              Assumed                SAR Assumed Power Bottom of the Reactor Core           Power Fraction. Fz                 Fraction
* r                     0.025                            0.5                           .3976 l                     0.075                            0.8                          1.2922 0.125                             1.0                         1.5712 6.175                             1.1                         1.5935 0.225                             1.1                          1.497 0.275                             1.1                         1.3671 l                     0.325                             1.1                         1.2399 0.375                             1.1                         1.1268 l                     0.425                             1.1                         1.0299 1                     0.475                             1.1                         .94865 0.525                             1.1                         .88209 0.575                             1.1                         .83005 l
0.625                             1.1                         .79305
[
!-                    0.675                             1.1                         .771 %
0.725                             1.1                         .76714 O.775                             1.1                         .77663 0.825                             1.1                         .79266 0.875                             1.0                         .79566 0.925                             0.8                         .74645 0.975                             0.5                         .59113 m This axial profile was used in the Cycle 1 analysis. A similar profile was used in Cycle 2; however, the information was not included in the SAR.
i

AM*A mmat1to 2CAN029807 Page 19 A flat power distribution has been used in the amendment request CEA ejection analyses.
This shape will cause the maximum amount of pin length to be operating at or near the peak power thus mW= Mag the moderator temperature. The increased moderator temperature will maximize the fuel temperature. Sensitivity studies have shown this shape to produce more conservative results.
Table 31 Assumptions For The Loss Of Load To One Steam Generator Amendment Parameter                     Units       Reauest Assumed      SAR Assumed Value Value l
l    Initial Core Power Level                  (MWt)              2534                   2900 RCS Pressure                               (psia)            2250                   2200 l
Doppler Multiplier                          -               0.85                    1.15 l
The results of this analysis are used to calculate a required overpower margin (ROPM) which,
[
l for this event, is dependent on the relative change in the power distribution from the beginning
'    of the event to the time of minimum DNBR. The reduced initial power level assumed in the amendment request analysis extends the duration of the transient by slowing the rate of increase of the cold leg temperature differential between the affected and unaffected steam generators, which extends the time to reach the CPC asymmetric steam generator trip. This allows more time for the distortion of the power distribution, producing a larger relative difference from the start of the event to the time of minimum DNBR. The resulting ROPM is therefore larger and more conservative.
The analysis results are insensitive to the initial RCS pressure, so a nominal value was assumed for the amendment request analyses.
A BOL doppler coefficient and multiplier were used in the amendment request analysis. The analysis results are not highly sensitive to this parameter, but the minimum doppler feedback  i minimizes the rate of power decay after the trip.                                              l I
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l Attachment 4 l       The following attachment contains the affidavit from Combustion Engineering, Inc. stating that the requested copy of Revision 2 of the ANO-2 Cycle 12 HZP MSLB Dose Evaluation contained in this letter is considered proprietary by Combustion Engineering,Inc.
l l
l

AFFIDAVIT PURSUANT TO 10 CFR 2.790 l, l.C. Rickard, depose and say that I am the Director, Operations Licensing, of Combustion Engineering, Inc., duly authorized to make this affidavit, and have reviewed or caused to have reviewed the information which is identified as proprietary and referenced in the paragraph immediately below. I am submitting this affidavit in conjunction with the application of Entergy, Inc., and in conformance with the provisions of 10 CFR 2.790 of the Commission's regulations.
The information for which proprietary treatment is sought is contained in the following document:
A-AN-FE-0132/95-E-0076-05, Rev. 02, "ANO-2 Cycle 12 HZP MSl_B Dose Evaluation," July 1997 This document has been appropriately designated as proprietary.
I have personal knowledge of the criteria and procedures utilized by Combustion Engineering in designating information as a trade secret, privileged or as confidential commercial or financial information.
Pursuant to the provisions of paragraph (b) (4) of Section 2.790 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure, included in the above referenced document, should be withheld.
: 1. The information sought to be withheld from public disclosure, is owned and has been held in confidence by Combustion Engineering. It consists of information concerning Combustion Engineering's methods for calculating      ;
radiological doses,                                                         j J
: 2. The information consists of test data or other similar data conceming a process, method or component, the application of which results in substantial competitive advantage to Combustion Engineering.
: 3. The information is of a type customarily held in confidence by combustion Engineering and not customarily disclosed to the public. Combustion Engineering has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certin types of information in confidence. The details of the aforementioned system were provided to the Nuclear Regulatory Commission via letter DP-537 from F. M. Stem to Frank Schroeder dated December 2,1974. This system was applied in determining that the subject document herein is proprietary.
: 4. The information is being transmitted to the Commission in confidence under the provisions of 10 CFR 2.790 with the understanding that it is to be received in confidence by the Commission.
: 5. The information, to the best of my knowledge and belief, is net available in public sources, and any disclosure to third parties has been made pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence.
: 6. Public disclosure of the information is likely to cause substantial harm to the competitive position of Combustion Engineering because:                         ,
I
: a. A similar product is manufactured and sold by major pressurized water reactor competitors of Combustion              !
Engineering.
: b. Development of this information by Combustion Engineering required tens of thousands of dollars and hundreds of

manhours of effort. A competitor would have to undergo similar expense in generating equivalent information.
: c. In order to acquire such information, a competitor would also require considerable time and inconvenience to develop similar methods for calculating radiological doses,
: d. The information consists of Combustion Engineering's methods for calculating radiological doses, the application of which provides a competitive economic advantage. The availability of such information to competitors would enable them to modify their product to better compete with Combustion Engineering, take marketing or other actions to improve their product's position or impair the position of Combustion Engineering's product, and avoid devr6 ping similar data and analyses in support of their processes,       l methods or apparatus.
: e. In pricing Combustion Engineering's products and services, significant research, development, engineering, analytical,    j manufacturing, licensing, quality assurance and other costs and expenses must be included. The ability of Combustion Engineering's competitors to utilize such information without similar expenditure of resources may enable them to sell at prices reflecting significantly lower costs.
: f. Use of the information by competitors in the international marketplace would increase their ability to market nuclear steam supply systems by reducing the costs associated with their technology development. In addition, disclosure would have an adverse economic impact on Combustion Engineering's potential for obtaining or maintaining foreign licensees.

I Further the deponent sayeth not.                        e i                                              Id"               /
ector Operations Licensing
!   Sworn to before me this  /Y         day of          4A4L-   .1998
  ,Y          -
y   -
Notary Public My commission expires:    3/ 97 l
4 l
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