ML050490153
| ML050490153 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 03/08/2005 |
| From: | Dan Collins NRC/NRR/DLPM/LPD1 |
| To: | Darrell Roberts NRC/NRR/DLPM/LPD1 |
| Collins D S, NRR/DLPM, 415-1427 | |
| References | |
| TAC MC3390 | |
| Download: ML050490153 (8) | |
Text
March 8, 2005 MEMORANDUM TO: Darrell J. Roberts, Chief, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation FROM:
Daniel S. Collins, Senior Project Manager, Section 2 /RA/
Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation
SUBJECT:
HOPE CREEK GENERATING STATION, FACSIMILE TRANSMISSION, DRAFT REQUEST FOR ADDITIONAL INFORMATION (RAI) TO BE DISCUSSED IN AN UPCOMING CONFERENCE CALL (TAC NO. MC3390)
The attached draft RAI was transmitted by facsimile on February 16, 2005, to Mr. Paul Duke, PSEG Nuclear, LLC (PSEG). This draft RAI is related to PSEGs submittal dated June 7, 2004, regarding an application requesting approval to allow an expanded operating domain resulting from implementation of the Average Power Range Monitor/Rod Block Monitor/Technical Specifications/Maximum Extended Load Line Limit Analysis (ARTS/MELLLA). Additionally, the proposed change would revise the methods used to evaluate annulus pressurization and jet loads resulting from the postulated Recirculation Suction Line Break. This draft RAI was transmitted to facilitate the technical review being conducted by the Nuclear Regulatory Commission (NRC) staff and to support a conference call with PSEG in order to clarify certain items in the licensees submittal. Review of the draft RAI would allow PSEG to determine and agree upon a schedule to respond to the RAI. This memorandum and the attachment do not convey a formal RIA or represent an NRC staff position.
Docket No. 50-354
Enclosure:
As stated
March 8, 2005 MEMORANDUM TO: Darrell J. Roberts, Chief, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation FROM:
Daniel S. Collins, Senior Project Manager, Section 2
/RA/
Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation
SUBJECT:
HOPE CREEK GENERATING STATION, FACSIMILE TRANSMISSION, DRAFT REQUEST FOR ADDITIONAL INFORMATION (RAI) TO BE DISCUSSED IN AN UPCOMING CONFERENCE CALL (TAC NO. MC3390)
The attached draft RAI was transmitted by facsimile on February 16, 2005, to Mr. Paul Duke, PSEG Nuclear, LLC (PSEG). This draft RAI is related to PSEGs submittal dated June 7, 2004, regarding an application requesting approval to allow an expanded operating domain resulting from implementation of the Average Power Range Monitor/Rod Block Monitor/Technical Specifications/Maximum Extended Load Line Limit Analysis (ARTS/MELLLA). Additionally, the proposed change would revise the methods used to evaluate annulus pressurization and jet loads resulting from the postulated Recirculation Suction Line Break. This draft RAI was transmitted to facilitate the technical review being conducted by the Nuclear Regulatory Commission (NRC) staff and to support a conference call with PSEG in order to clarify certain items in the licensees submittal. Review of the draft RAI would allow PSEG to determine and agree upon a schedule to respond to the RAI. This memorandum and the attachment do not convey a formal RIA or represent an NRC staff position.
Docket No. 50-354
Enclosure:
As stated DISTRIBUTION:
PUBLIC DRoberts RDennig ECoby, RGN-I ACRS DCollins EMarinos CRaynor OGC PDI-2 Reading EThrom RLobel GMiller EKendrick MRazzaque PRebstock ZAbdullahi FAkstulewicz ADAMS Accession Number: ML050
- Questions provided by Memorandum OFFICE PDI-2/PE PDI-2/PM SRXB-A/SC SPSB-C/SC NAME GMiller DCollins FAkstulewicz RDennig*
DATE 3/7/05 3/7/05 2/15/2005 10/13/2004 OFFICIAL RECORD COPY
DRAFT REQUEST FOR ADDITIONAL INFORMATION REGARDING AMENDMENT REQUEST TO UTILIZE ARTS/MELLLA OPERATING DOMAIN HOPE CREEK GENERATING STATION DOCKET NO. 50-354 By letter dated June 7, 2004, PSEG Nuclear, LLC (PSEG) submitted license amendment request LCR H04-01 for the Hope Creek Generating Station (Hope Creek), seeking U.S.
Nuclear Regulatory Commission approval of proposed changes to the Hope Creek Technical Specifications (TSs). Specifically, the proposed change would allow an expanded operating domain resulting from implementation of Average Power Range Monitor/Rod Block Monitor/Technical Specifications/Maximum Extended Load Line Limit Analysis (ARTS/MELLLA). Additionally, the proposed change would revise the methods used to evaluate annulus pressurization and jet loads resulting from the postulated Recirculation Suction Line Break. The NRC has developed the following draft questions during its review of the application. For clarity, the questions have been grouped by the various sections they came from.
1.
Containment and Accident Dose Assessment Section a)
The current licensed thermal power (CLTP) is 3339 MWt. The original licensed thermal power (OLTP) was 3293 MWt. The referenced updated final safety analysis report analyses were performed at 102% of the OLTP, which suggests that the containment studies would be equivalent to 100.6% of the CLTP.
Clarify this apparent discrepancy or provide justification for reducing the conservatism in the containment performance studies from the previous value of 102% of the nominal operating power to 100.6% of the nominal operating power.
b)
In your submittal, you reference General Electric Co. (GE) Nuclear Energy, Technical Description - Annulus Pressurization Load Adequacy Evaluation, NEDO-24548, January 1979. Please provide a copy of this report to assist the staff in its review of the proposed reactor c)
Provide a discussion of the differences in the calculation of the subcompartment loads previously used for the reactor asymmetric loads evaluation (the COPDA code) to the proposed revised method using the COMPARE code. Include, for example, the treatment of heat structures, the treatment of the water-steam-air mixture for mass and heat transfer, and the treatment of the mass and energy as they enter the break control volume for mass and energy partitioning between the fluid and vapor states. Provide comparison graphs of the limiting control volume and differential pressures, over the time period of interest, for both methods.
Enclosure d)
Provide a design description of annulus region and the existing 25/75 flow diverter which limits the break flow into the annulus to 25% of the nominal value. Where is the postulated break located (i.e., on the recirculation or the feedwater line) and where does the remaining flow go? Is this flow reduction the major contributor to meeting the design requirements? Was the diverter considered in the original code calculations? If not, was a COPDA calculation does with the diverter, and what were the results?
e)
Provide a discussion of the differences in the calculation of the mass and energy releases previously used for the reactor asymmetric loads evaluation (NEDO-24548) to the proposed revised method using the code. Include, for example, the critical flow models used, the equations of state for water used, and the initial primary systems conditions (pressure, temperature, mass, etc.).
Provide comparison graphs of the mass and the energy releases, over the time period of interest, for both methods.
f)
Provided a discussion of the differences between the COPDA and the COMPARE models for the reactor annulus, including nodalization and the treatment of inter-compartment flow paths. Describe the COMPARE sensitivity studies (nodalization, flow paths, time step, etc.) Performed to develop the final model for use in licensing analyses.
g)
Provide a discussion addressing quality assurance control, as delineated in Title 10 of the Code of Federal Regulations Part 50, Appendix B, for the COMPARE analyses that are performed to support licensing actions. Include, for example, configuration control, user training and data validation and verification for model development.
h)
On page 10 of your submittal you state that the GENE LAMB code (NEDE-20566P-A) will be used to provide a more realistic blowdown mass and energy release profile. This would be in lieu of the original NEDO-24548 methodology.
Is the COMPARE code part of the NEDO-24548 methodology? What methods are used by GE for the AP pressure time history conversion to nodal force time history? What structural model is used?
2.
Instrumentation and Control Section a)
The Hope Creek technical specifications define Limiting Safety System Settings (LSSS) as an allowable value (AV). During reviews of proposed license amendments that contain changes to LSSS setpoints, the NRC staff identified concerns regarding the method used by some licensees to determine the allowable values (AV) identified in the technical specifications (TS). AVs are identified in the TS as LSSS to provide acceptance criteria for determination of instrument channel operability during periodic surveillance testing. The NRC staffs concern relates to one of the three methods for determining the AV as described in the Instrument Society of America (ISA) recommended practice ISA-RP67.04-1994, Part II, Methodologies for Determination of Setpoints for Nuclear Safety-Related Instrumentation.
The staff has determined that, absent additional requirements related to determining the operability of the instrument channel, AVs associated with LSSS established by means of ISA-RP67.04, Part II, Method 3, will not provide reasonable assurance that a plant will operate in accordance with the assumptions upon which the plant safety analyses have been based. Details about the NRC staff's concerns are available on the NRC's public website under ADAMS Accession Numbers ML041690604 and ML041810346.
In Order for the NRC staff to assess the acceptability of your license amendment request related to this issue, the NRC staff requests the following additional information:
1.
Discuss the setpoint methodology used at Hope Creek to establish AVs associated with LSSS setpoints.
2.
Regardless of the methodology used, the NRC staff has the following questions regarding the use of the methodology at [insert plant name]:
a.
Discuss how the methodology and controls you have in place ensure that the analytical limit associated with an LSSS trip setpoint will not be exceeded (that safety limits will not be exceeded). Include in your discussion information on the controls you employ to ensure the trip setpoint established after completing periodic surveillances satisfies your methodology. If the controls are located in a document other than the TS, discuss how those controls satisfy the requirements of 10 CFR 50.36.
b.
Discuss how the TS surveillances ensure the operability of the instrument channel. Specifically, relate the surveillance test results to the technical specification AV and describe how these are used to determine the operability of the instrument channel. If the requirements for determining operability of the LSSS instrument being tested are in a document other than the TS (e.g.,
plant test procedure), discuss how this meets the requirements of 10 CFR 50.36.
b)
In your submittal, you state that the APRM flow control trip reference (FCTR) cards will be modified to implement the proposed setpoint changes. Please elaborate on this modification. Are the FCTR cards analog or digital, and if digital, will the modifications be accomplished by software or firmware changes?
3.
Boiling Water Reactor and Nuclear Performance Section a)
What was the HCGS original thermal analysis basis, and was it changes before or during the change to the current licensing basis using the Extended load line limit analysis (ELLLA)/ICA extended power/flow map for Cycle 2 (Reference 2 of Attachment 1 of your June 7, 2004 submittal). Please cite references to GETAB and GESTAR_II as appropriate.
b)
How does the implementation of the proposed partial ARTS improvement interact with and complement the proposed MELLLA implementation to replace the current ELLLA power flow map. Please confirm that the current ICF region is not affected. Figures showing the current ELLLA versus the proposed MELLLA regions and the current APRM/RBM flow-biased setpoints versus the proposed setpoints would be helpful.
c)
In the HCGS MSAR (NEDC-33066P), a representative core with all GE14 fuel is used to provide representative analyses results at the proposed ARTS/MELLLA conditions. Provide a description of this representative core. Is this representative core based on a projected equilibrium loading for HCGS, a generic plant similar to HCGS, or a loading pattern considered bounding for the HCGS at the CLTP?
d)
Will the SRLR (and COLR) for Cycle 13 address both ELLLA and ARTS/MELLLA conditions for the actual mixed fuel core loading?
e)
Your submittal states that the rod block monitor (RBM) is not credited in the Rod Withdrawal Error analysis and is described to be a power generation system, not used for accident mitigation.
i.
Was the RBM credited in the original design of the HCGS? If so, provide a reference for this change.
ii.
Is the RBM being deleted from the plant design f)
The current TS Limiting Condition for Operation (LCO) 3.2.2 requires the APRM flow-biased scram and rod block trip setpoints to be reduced (setdown) when the fraction of rated thermal power (FRTP) is less than the core maximum fraction of limiting power density (CMFLPD). It is stated that this APRM setdown requirement originated from the Hench-Levy Minimum Critical Heat Flux Ratio (MCHFR) thermal limit criterion. Provide a reference for the Hench-Levy MCFHR thermal limit criterion.
g)
In your submittal, you discuss a flow diverter that was added prior to initial HCGS plant operation. This flow diverter was added to the shield wall penetration to mitigate the AP load effects, by reducing the fraction of blowdown flow entering the annulus to 25%. The addition of the flow diverter created a jet load on the diverter/shield wall which did not exist in the original design basis. The installation of the diverter was evaluated by a qualitative assessment of the effects on the original analysis, however, the combined effects of the new set of loads were never quantified in detail. Describe, in detail, the combined effects of the new set of loads.
h)
In your submittal you state that the peak vessel pressure, the suppression pool temperatures, pool level, and the corresponding containment pressures were obtained using approved methods to evaluate the limiting ATWS scenario.
Provide a reference to the specific approved methods used.
i)
Section 11.0 of Attachment 4 of your submittal discusses the effects of the increased subcooling due to MELLLA operation on the current HCGS high energy line break (HELB) design bases. The results of the GE analyses show that the current mass and energy release profiles are bounding for all HELB locations except for the four break locations for the Reactor Water Cleanup (RWCU) system line. It is stated that PSEG has evaluated the effect of the higher mass and energy release profiles for the RWCU and has concluded that the resulting subcompartment pressures, temperatures, and humidity levels are acceptable with respect to the existing design criteria. Elaborate on the PSEG evaluation and state the design criteria used.
j)
As discussion of the HCGS currently licensed APRM, RBM, and ELLLA operating map restrictions is provided in Section 1.1. Elaborate on the proposed partial implementation of the ARTS improvements in conjunction with the proposed MELLLA operating map. Specifically, why is only a partial ARTS implementation required and how does it complement the MELLLA mode?
k)
Provide a graphical representation of the current ELLLA/ICF domain.
l)
In your submittal you state that the cycle-specific analyses will be performed by PSEG using approved methodologies for HCGS. Provide a list of the approved methodologies to be used.
m)
The submittal states that the APRM flow-biased simulated thermal power scram line and the APRM flow-biased rod block line are not credited in any HCGS safety licensing analyses. However, since they remain in the HCGS design configuration and TS, it is proposed to change the setpoints for operational flexibility. What controls will be used to prevent unintended operation above the MELLLA line given that operators will no longer be able to rely on APRM setpoint alarms to warn of this condition.
n)
Section 1.2.1 of Attachment 4 to your submittal states that PSEG will perform the actual cycle-specific analyses for the reload licensing activity using approved HCGS methodologies. Are the methodologies referred to based on the GESTAR methods or on the currently approved HCGS methods?
o)
In Section 3.0 of Attachment 4 to your submittal, six events are listed as potentially limiting for OLMCPR and were analyzed or reviewed for HCGS operation in the MELLLA region. No results are shown for the TTNBP event, but it is not listed as non-limiting. Was the TTNBP event analyzed or reviewed.
If so, was it found to be non-limiting.
p)
LFWH and inadvertent HPCI were not analyzed at the MELLLA conditions.
What is the basis for the conclusion that there is a large margin to other events? Since LFWH/HCPI can occur at normal power operations, why is this event not considered in the determination and validation of the off-rated limits?
q)
In Section 3.0 of Attachment 4 to your submittal, you state that IRLS and FRFI events have been considered generically for the introduction of GE-14 fuel and are bounded by the generic ARTS/MELLLA power and flow dependent limits.
Confirm that the anticipated operational occurrences (AOO) analyses conclusions are based on actual HCGS plant specific ARTS/MELLLA evaluations and analyses. If not, please discuss why this is not necessary.
r)
In Section 10.1 of Attachment 4 to your submittal, you state that MSIVC and PRFO are the two limiting events considered and that the PRFO event at 120%
of OLTP and 99% of RCF is the limiting event for HCGS.
It appears that loss of offsite power and IROV were not considered. How were they determined to be non-limiting?