2CAN070204, Arkansas Unit 2 Supplement to Amendment Request to Revise Section 6.0, Administrative Controls
| ML022070511 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 07/18/2002 |
| From: | Cotton S Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 2CAN070204 | |
| Download: ML022070511 (7) | |
Text
Entergy Operations. Inc.
Entergy 1448 S R 333 Russellvl le AR 72802 Tel 50' 858 5109C 2CAN070204 July 18, 2002 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
SUBJECT:
Arkansas Nuclear One, Unit 2 Docket No. 50-368 Supplement to Amendment Request to Revise Section 6.0, Administrative Controls
REFERENCES:
- 1. Letter dated June 26, 2002, Revision of Section 6.0, Administrative Controls For Consistency with ANO-1 Improved Technical Specifications (2CAN060203)
- 2. Letter dated January 31, 2002, Revision of Section 6.0, Administrative Controls For Consistency with ANO-1 Improved Technical Specifications (2CAN010203)
Dear Sir or Madam:
By letters (reference 1 and 2) Entergy Operations, Inc. (Entergy) proposed a change to the Arkansas Nuclear One, Unit 2 (ANO-2) Technical Specifications (TSs) to reorganize the Administrative Controls Section of the ANO-2 TSs, which also included changes to several other TSs.
The purpose of this letter is to correct some electronic errors, as well as others that were identified upon a subsequent review, which occurred in the transmittal of the letter dated June 26, 2002 (reference 1). The corrected pages are included in Attachment 1 to this letter and should replace the existing pages or be inserted in the appropriate attachment of the letter.
In the cover letter of 2CAN060203 (reference 1), the date of the letter referenced as reference #1 incorrectly stated the date of the letter as being January 23, 2002. The actual date of the letter was January 31, 2002.
The following correction will be made and will be designated with a revision bar by the change:
In Attachment 1 to 2CAN060203 (reference 1), on page 52 of 56, portions of the information contained in paragraph 36.0 are missing. The following should have been included in the letter:
2CAN070204 Page 2 of 3 "36.0 New 6.5.14, Technical Specification (TS) Bases Control Program A TS bases control program will be added to the ANO-2 TS. This program is necessary for proper implementation of the conversion and is consistent with the ANO-1 ITS and meets the intent of NUREG-1432.
The program addition is an additional restriction on unit operation."
On page B 3/4 9-3 in Attachment 3 to 2CAN060203, Revised Markup of Technical Specification Bases Pages, the header "3/4 8 Electrical Power Systems, Bases" was inappropriately carried over to this page from the previous page. The header will be deleted. There were two pages associated with page B 3/4 9-3 included in 2CAN060203 due to the additional space taken by the inappropriate header. When the header is removed, the markup results in only one page. No revision bars will be included for this change.
In Attachment 5 to 2CAN060203, Revised Clean Technical Specification Bases Pages, the clean bases page associated with the proposed changes to specification 3/4.9.11 were inadvertently omitted from the letter. The appropriate revision bars will be included on this page.
There are no technical changes proposed. The original no significant hazards considerations included in reference 1 is not affected by any information contained in this supplemental letter. There are no new commitments contained in this letter.
If you have any questions or require additional information, please contact Dana Millar at 601-368-5445.
I declare under penalty of perjury that the foregoing is true and correct. Executed on July 18, 2002.
Sincerely, Sherrie R. Cotton Director Nuclear Safety Assurance SRC/dm Attachments:
- 1. Corrected Pages
2CAN070204 Page 3 of 3 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. 0. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. Thomas W. Alexion MS O-7D1 Washington, DC 20555-0001 Mr. Bernard R. Bevill Director Division of Radiation Control and Emergency Management Arkansas Department of Health 4815 West Markham Street Little Rock, AR 72205
Attachment 1 To 2CAN070204 Corrected Pages to 2CAN060203 Page 52 of 56
" The proposed change does not replace the word "containment" with "reactor building." The title Containment Leakage Rate Testing Program and reference to "containment" instead of "reactor building" is consistent with NUREG-1432. This does not present a change to the current wording contained in the ANO-2 TSs.
"* The peak calculated containment internal pressure for the design basis loss of coolant accident for ANO-2 is 58 psig.
"* The ANO-2 CTS requires that the maximum allowable containment leakage rate, La, shall be 0.1% of containment air weight per day at Pa. This is the current ANO-2 licensing basis and no change is proposed.
"* Air lock acceptance criteria is included in the ANO-2 CTS. This is consistent with the currently approved ANO-2 TS and no change is proposed.
NUREG-1432 Comparison Minor wording differences exist between the ANO-2 proposed TS and NUREG 1432. These differences, however, do not modify the intent of the words contained in NUREG-1432.
NUREG-1432 specification 5.5.16 [OPTION B] does not include the testing requirements related to the containment purge supply and exhaust isolation valves. Relocation of this requirement from CTS 4.6.3.1.4 does not change the intent of the NUREG section. It consolidates the testing requirements in one location.
NUREG-1432 specifies Type A leakage rate acceptance criteria of *0.75 L,. The ANO-1 ITS modified this acceptance criteria as described above. The ANO-2 proposed TS is consistent with the ANO-1 ITS and thus differs from NUREG 1432. This is justified above.
NUREG-1432 states, "Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J."
The ANO-2 CTS states, "The provisions of Specification 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program."
The CTS words, which are consistent with the wording in the ANO-1 ITS, reflect the intent of NUREG-1432. No change is proposed to the CTS words.
36.0 New 6.5.14, Technical Specification (TS) Bases Control Program A TS bases control program will be added to the ANO-2 TS. This program is necessary for proper implementation of the conversion and is consistent with the ANO-1 ITS and meets the intent of NUREG-1432. The program addition is an additional restriction on unit operation.
REFUELING OPERATIONS BASES 3/4.9.9 and 3/4.9.10 WATER LEVEL-REACTOR VESSEL AND SPENT FUEL POOL WATER LEVEL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 12% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the accident analysis.
3/4.9.11 FUEL HANDLING AREA VENTILATION SYSTEM The limitations on the fuel handling area ventilation system ensure that all radioactive materials released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorbers prior to discharge to the atmosphere. The operation of this system and the resulting iodine removal capacity are consistent with the assumptions of the accident analyses.
AGceptable removal etcencY is shown by methyh iodide penetration of less than 5.0% when tests are..performed.in.accordance with.ASTM.£D3803-1989. "StandardTestMethod-for Nuclear-Grade Activated Carbon,*"'*at.a-temperature of 309Cand-.a-relative humidity-of 95%.
The penetration aceptance criterion is determ.ned by the following equation:
Allowable-Oeý-[0 --methy-iodide-efficieny-or-haarcoal-credftte-4n -aGcidt-analysis]
Pene tration ... . ... ......... ........ ........ .... .. .safety. facto r of -2 Applying-a safety factor of 2 is acceptable-ber-ause ASTMV D3803 1989 is a more-accurate and.demanding test than-older-tests.
3/4.9.12 FUEL STORAGE Region 1 and Region 2 of the spent fuel storage racks are designed to assure fuel assemblies of less than or equal to 5.0 w/o U-235 enrichment that are within the limits of Figure 3.9.2 will be maintained in a subcritical array with Keff *<0.95 in unborated water.
These conditions have been verified by criticality analyses.
The requirement for 1600 ppm boron concentration is to assure the fuel assemblies will be maintained in a subcritical array with Keff *0.95 in the event of a postulated accident. Analysis has shown that, during a postulated accident with the fuel stored within the limits of this specification, that a Keff of *0.95 will be maintained when the boron concentration is at or above 1000 ppm.
Normally, fuel stored in a cross-hatch storage configuration must have all four diagonal spaces or at least two adjacent faces remain vacant to meet the criticality safety analysis mentioned above. However, the spent fuel pool walls may be credited as a neutron leakage path. Therefore, vacant spaces face adjacent to the walls of the Region I cross-hatch configured assemblies may be used to store fuel assemblies that are outside of the area of the graph enclosed by Curve A on Figure 3.9.2, excluding the most southeast and southwest corner spaces of Region 1 which must remain empty.
ARKANSAS - UNIT 2 B 3/4 9-3 Amendment No. 43,466,-78,224,2-28,
REFUELING OPERATIONS BASES 3/4.9.9 and 3/4.9.10 WATER LEVEL-REACTOR VESSEL AND SPENT FUEL POOL WATER LEVEL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 12% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the accident analysis.
3/4.9.11 FUEL HANDLING AREA VENTILATION SYSTEM The limitations on the fuel handling area ventilation system ensure that all radioactive materials released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorbers prior to discharge to the atmosphere. The operation of this system and the resulting iodine removal capacity are consistent with the assumptions of the accident analyses.
3/4.9.12 FUEL STORAGE Region 1 and Region 2 of the spent fuel storage racks are designed to assure fuel assemblies of less than or equal to 5.0 w/o U-235 enrichment that are within the limits of Figure 3.9.2 will be maintained in a subcritical array with Keff *0.95 in unborated water.
These conditions have been verified by criticality analyses.
The requirement for 1600 ppm boron concentration is to assure the fuel assemblies will be maintained in a subcritical array with Keff *0.95 in the event of a postulated accident. Analysis has shown that, during a postulated accident with the fuel stored within the limits of this specification, that a Kef of *0.95 will be maintained when the boron concentration is at or above 1000 ppm.
Normally, fuel stored in a cross-hatch storage configuration must have all four diagonal spaces or at least two adjacent faces remain vacant to meet the criticality safety analysis mentioned above. However, the spent fuel pool walls may be credited as a neutron leakage path. Therefore, vacant spaces face adjacent to the walls of the Region I cross-hatch configured assemblies may be used to store fuel assemblies that are outside of the area of the graph enclosed by Curve A on Figure 3.9.2, excluding the most southeast and southwest corner spaces of Region 1 which must remain empty.
ARKANSAS - UNIT 2 B 3/4 9-3 Amendment No. 43,1-66,-178,224,228,