View source for HL-5362, Application for Amends to Licenses DPR-57 & NPF-5,revising Operability Requirements for Rod Block Monitor Sys

{{Adams
| number = ML20141B835
| issue date = 05/09/1997
| title = Application for Amends to Licenses DPR-57 & NPF-5,revising Operability Requirements for Rod Block Monitor Sys
| author name = Sumner L
| author affiliation = SOUTHERN NUCLEAR OPERATING CO.
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000321, 05000366
| license number = 
| contact person = 
| document report number = HL-5362, NUDOCS 9705160041
| package number = ML20141B837
| document type = OPERATING LICENSES-APPLIATION TO AMEND-RENEW EXISTING, TEXT-LICENSE APPLICATIONS &amp; PERMITS
| page count = 7
| project = 
| stage = Request
}}

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s *e Lewis Sumner               Southern Nuclear Vice President             Operating Company, loc.
Hatch Project Support      40 invemess Parkway
       '*A                                    Post Office Box 1295 Birmingham, Alabama 35201 Tal 205.992.7279 Fax 205.992.0341 SOUTHERN L COMPANY Energy to Serve hurWorld" May 9, 1997 Docket Nos. 50-321                                                                              HL-5362 50-366 U.S. Nuclear Regulatory Conunission ATTN: Document Control Desk Washington, D.C. 20555 Edwin I. Hatch Nuclear Plant Request to Revise Technical Specifications:
RcxlBlock Monitor Operability Reauirements
             ' Gentlemen:
In accordance with the provisions of 10 CFR 50.90, as required by 10 CFR 50.59(c)(1),
Southern Nuclear Operating Company (SNC) hereby proposes a change to the Plant Hatch Units 1 and 2 Technical Specifications, Appendix A to Operating Licenses DPR-57 and NPF-5, respectively. The proposed changes revise the operability requirements for the Rod Block Monitor system.
Enclosure 1 provides a description of the proposed changes and an explanation of the basis for the changes. Enclosure 2 details the basis for SNC's determination that the proposed changes do not involve a significant hazards consideration. Enclosure 3 provides page change instructions for incorporating the proposed changes into the                              /
Technical Specifications. Following Enclosure 3 are the revised Technical Specifications                      /
pages and the corresponding marked up pages. Enclosure 4 provides page change                                      )
instructions for incorporating the proposed changes into the Bases. Following                                   '/
Enclosure 4 are the revised Bases pages and the corresponding marked up pages.
                                                                                                                                     /
Southern Nuclear Operating Company requests the proposed amendments be issued as                                     f soon as possible, with an immediate effective date and implementation within 30 after issuance for Unit 2 and Unit I to be effective and implemented prior to Unit I startup from the Fall of 1997 refueling outage.
In accordance with the requirements of 10 CFR 50.91, the designated State ofFcial will be sent a copy of this letter and all applicable enclosures.
9705160041 970509 PDR      ADOCK 05000321 I
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* I U.S. Nuclear Regulatory Commission                                                             Page 2 Mr. H. L. Sumner states he is Vice President of Southern Nuclear Operating Company and is authorized to execute this oath on behalf of Southern Nuclear Operating Company, and to the best of his knowledge and belief, the facts set forth in this letter are true.
Sincerely, l
[f4}O  /        [M H. L. Sumner, Jr.
vfl Sworn to andsubscribed before me this                        dayof  &     W&                     1997.         \
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NotaryPublic j Commission Expiration Date:                      //-2-9S l
WRM

==Enclosures:==
: 1. Basis for Change Request
: 2. 10 CFR 50.92 Evaluation
: 3. Page Change Instructions
: 4. Bases Changes cc: Southern Nuclear Operatine Company Mr. P. H. Wells, Nuclear Plant General Manager NORMS U.S. Nuclear Reerdatorv Commission. Washington. D.C.
Mr. K. Jabbour, Licensing Project Manager - Hatch U.S. Nuclear Regulatorv Commission. Region 11 Mr. L. A. Reyes, Regional Administrator Mr. B. L. Holbrook, Senior Resident Inspector - Hatch
                    ' State of Georgia
:                    Mr. J. D. Tanner, Commissioner - Department of Natural Resources HL-5362 l                                                                                                     ..

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* Enclosure 1 Edwin I. Hatch Nuclear Plant Request to Revise Technical Specifications:
Rod Block Monitor Operability Requirements pasis for Chance Reauest Proposed Chance The proposed amendments modify the Units 1 and 2 Technical Specifications Tables 3.3.2.1-1 to revise the operability requirements for the control Rod Block Monitor (RBM) system. The changes require the RBM to be operable at all times when core thermal power is equal to or greater than 29% of rated thermal power. These modifications are more restrictive than the current Technical Specifications requirements. The amendments also delete the requirements in Technical Specifications sections 5.6.5 to report Rod Block Monitor operability requirements in the cycle-specific Core Operating Limits Report.
Basis for Proposed Change in December,1984, the NRC approved the installation of the Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvements (ARTS) program for Edwin I. Hatch Nuclear Plant Units 1 and 2 (References 1 and 2). The ARTS hardware modifications included a reconfiguration of the RBM system to prevent violation of the Minimum Critical Power Ratio Safety Limit (SLMCPR) during a control      I rod withdrawal error (RWE) when the reactor is operating at power. The Plant Hatch Unit I and Unit 2 Technical Specifications were also amended to specify new operability requirements for the modified RBM system. Those requirements were the result of a generic ARTS analysis for Plant Hatch (Reference 3) which determined the plant conditions under which the complete withdrawal of a deep control rod could result in a violation of the SLMCPR. The function of the RBM is to limit rod withdrawal distance when there is a significant increase in the local power around a rod that is being withdrawn. To prevent violations of the SLMCPR, rod blocks may be necessary whenever the core average thermal power is equal to or greater than 29% of rated and the operating MCPR is below 1.40 or 1.70, depending on the core thermal power. Thus, for example, when core thermal power is greater than 90% of rated and the MCPR operating value is less than 1.40, the Technical Specifications require both channels of the RBM to be operable.
A control rod block from the RBM can also prevent fuel cladding damage due to mechanical overpowers during a rod withdrawal error event. As described in GE's j         licensing topical report for BWR fuel, GESTAR-II (Reference 4), fuel cladding damage is not expected to occur if the cladding strain does not exceed 1% during anticipated operational occurrences (AOO). For reload cores, compliance with this specified HL-5362                                      El-1

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Enclosure 1 Request to Revise Technical Specifications-Rod Block Monitor Operability Requirements .
Basis for Change Request l
acceptable fuel design limit (SAFDL) is evaluated for all transient events, including RWE, as part of the cycle-specific reload licensing analysis.
j          Recently, GE informed us that, for ARTS plants, their technical design procedure for
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evaluating fuel rod mechanical overpowers during an RWE contained a step which allows l
them to assume the RBM is operable, regardless of the MCPR operating value.                      '
According to the procedure, if the criteria for showing compliance with the mechanical overpower SAFDL cannot be met with the RBM inoperable, the analysis is re-run assuming RBM is operable. If the criteria is met under these conditions, it is concluded that the fuel cladding will not exceed 1% plastic strain during an RWE. This conclusion,         i however, may not always be valid since the Technical Specifications allow RBM to be bypassed under certain conditions (e.g., when core thermal power is greater than 90% of rated and MCPR is greater than 1.40). Thus, there is an inconsistency between the Technical Specifications requirements and the analysis assumption regarding the
:         operability of the RBM system.
In order to ensure that no fuel rod will exceed 1% plastic strain during an RWE event at power, Table 3.3.2.1-1 of the Technical Specifications, which contains the operability requirements for RBM system, is being revised to delete the provisions for bypassing RBM if the MCPR operating value exceeds certain power-dependent limits.
Subsequent to installation of ARTS at Plant Hatch, the NRC approved replacement of cycle-specific power distribution limits in the Technical Specifications with a reference to a Core Operating Limits Report (Reference 5). At that time the MCPR requirements for Rod Block Monitor operability were removed from the Technical Specifications and l         placed in the Section 2.0 of the COLR. When the Improved Technical Specifications were l         implemented at Plant Hatch in 1995, the RBM requirements were added back into the                 ,
Technical Specifications, but were not removed from the COLR. Since these                        1 l
l requirements are redundant and RBM operability will not be affected by MCPR, Section 5.6.5 of the Technical Specifications is being revised to delete RBM operability j         requirements from the COLR.                                                                      .
References
: 1. NRC letter issuing Amendment No.105 to Facility Operating License No.
DPR-57 for Edwin I. Hatch Nuclear Plant Unit 1, R. A. Hermann to J. T.
:                     Beckham, December 31,1984.
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Enclosure 1                                                                           l Request to Revise Technical Specifications:
Rod Block Monitor Operability Requirements                                            i Basis for Change Request
: 2. NRC letter issuing Amendment No. 39 to Facility Operating License No. NPF-5 for Edwin I. Hatch Nuclear Plant Unit 2, G. W. Rivenbark to J. T. Beckham, July 31,1984.
: 3. NEDC-30474-P," General Electric BWR Licensing Report: Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvement (ARTS) Program for Edwin I. Hatch Nuclear Plant, Units 1 and 2," December 1983.
: 4. NEDE-24011-P-A-13," General Electric Standard Application for Reactor Fuel (GESTAR-II)," August 1996.                                                 1
: 5. NRC letter issuing Amendment Nos.168 and 106 to Facility Operating Licenses 1 DPR-57 and NPF-5 for Edwin I. Hatch Nuclear Plant Units 1 and 2, L. P.
Crocker to W. G Hariston, III, December 29,1989.
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HL-5362                                    El-3

  ,. i Enclosure 2 Edwin I. Hatch Nuclear Plant Request to Revise Technical Specifications:
Rod Block Monitor Operability Requirements 10 CFR 50.92 Evaluation In 10 CFR 50.92(c), the NRC provides the following standards to be used in determining the existence of a significant hazards consideration:
               ...a proposed amendment to an operating license for a facility licensed under 50.21(b) or 50.22 or for a testing facility involves no significant hazards consideration,if operation of the facility in accordance with the proposed amendment would not: (1)
Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.
Basis for no sieniReant hazards determination                                                l l
Southern Nuclear Operating Company has evaluated the proposed changes to the Plant Hatch Units 1 and 2 Technical Specifications in accordance with the criteria specified in    I 10 CFR 50.92 and has determined that they do not involve a significant hazards consideration because:
: 1. The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated since they are more restrictive than the existing requirements for operation of the plant. These changes provide assurance that the Rod Block Monitor system will remain operable when necessary to prevent or mitigate the consequences of an anticipated operational occurrence that could threaten the integrity of the fuel cladding integrity. Since changes in RBM operability requirements do not involve any physical or functional modifications in any plant system, structure or component, there will be no increase in the probability or consequences of any accident previously evaluated.
: 2. The proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated because they do not involve any changes in the plant configuration or in the operation of any system, structure l                      or component.
: 3. The proposed changes do not reduce a margin of safety in the plant because they impose more restrictive operability requirements on the Rod Block Monitor system than those imposed by the existing specifications. The l
i HL-5362                                       E2-1 l

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,                              Enclosure 2                                                                                           )
l                    .       JRequest to Revise Technical Specifications:
l Rod Block Monitor Operability Requirements                                                          ,
i                              10 CFR 50.92 Evaluation changes are more restrictive in that they delete the conditions under which the RBM is              >
l                              allowed to be bypassed at core thermal power equal to or greater than 29% of rated power. These more restrictive requirements ensure the RBM will not only prevent fuel l                              rods from under going transition boiling, they also prevent fuel rods from exceeding 1%             i
                             - plastic strain (thereby avoiding fuel cladding damage) during an RWE event.
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