ML20141N814
| ML20141N814 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 03/13/1986 |
| From: | Tucker H DUKE POWER CO. |
| To: | Harold Denton, Youngblood B Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8603180159 | |
| Download: ML20141N814 (6) | |
Text
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< swasma PBestfT90N March 13, 1986 Mr. Narold R. Denton, Director Office of Nuclear Reactor Regulation
- U. 5. Ihnelaar Regulatory Commission Washington, D. C.
20555 Attentiont-Mr. B. J. Toungblood, Project Director FWR Project Directorate No. 4
Subject:
Catawba Nuclear Station, Unit 2 Docket No. 50-414
Dear Mr. Denton:
This letter contains a description of a change in the Initial Startup Test Program which.is reportable under License Condition 3 of Facility Operating License NFF-48 and 10 CFR 50.59(b). Attachment 1 is a copy of the marked-up FSAR page which will be incorporated into the next update of the Catawba FSAR.- Attachment 2 provides a copy of the evaluation conducted in
,e accordance with the requirements of 10 CFR 50.59.
n Prior to this change, a required step in the Core Verification Procedure use to verify proper Rod Cluster Control Assembly (RCCA) location by RCCA s'-
ID number. The revised procedure requires that a visual check be performed to verify that core locations desi~gnated for fuel assemblies with RCCA's have these assemblies in them. The check of RCCA locations with ID numbers is namecessary because this uns done as part of a previously completed procedure, the Initial Core Assembly Insert Verification.
In addition, a verification of proper RCCA location by ID number is difficult to conduct as well as time consuming.
Prior to commencing fuel loading, the Initial Core Assembly Insert Verification Procedure is completed. As a part of this procedure, the insertion of the proper RCCA's into the correct fuel assemblies is verified by RCCA and fuel assembly ID numbers. During and subsequent to fuel loading, the correct location and orientation of the fuel assemblies is verified.
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l Mr. Harald R. Derton, Direetcr March 13, 1986 Page 2 s
Since the Initial Core Insert Verification assures the correct RCCA-fuel assembly matches, and fuel assembly location is checked subsequent to fuel loading, there is reasonable assurance that the RCCA's will be in the correct locations.
In addition, the Core Verification Procedure requires a visual check to assure that there is an RCCA in each fuel assembly designated for one.
Therefore, it hr.s heen concluded that a verification of RCCA positions by ID numbers is unnecessary.
Very truly yours, b-lh)
Hal B. Tucker WLH: nib Attachments cc:
Dr. J. Nelson Grace, Regional Administrator U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 290C Atlanta, Georgia 30323 NRC Resident Inspector Catawba Nuclear Statfor k
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ATTACHMENT 1 CNS Fuel assemblies and inserted components are received, inspected, and placed in storage in accordance with written, approved procedures.
Prior to commenc-ing fuel loading each assembly will be inspected to verify that it contains the proper inserted component and that the component is properly oriented.
At the time of fuel loading, they are placed in the reactor vessel one at a time according to a previously-established, approved, written sequence which was developed to provide reliable core monitoring with minimum possibility of core mechanical damage. The fuel loading procedure documents include tabular check sheets which prescribe and verify the successive movements of each fuel assembly and its specified inserts from its initial position in the storage racks to its final position in the core.
Checks are made of component serial numbers y
and types at various transfer points to guard against possible inadvertant ex-changes or substitutions of components; however, in the event that mechanical damage is sustained during fuel loading operations, to a fuel assembly of a type for which no spare is available onsite, an alternate core loading scheme, whose characteristics closely approximate those of the initial prescribed pat-tern, is determined and all physics parameters specified for the initial design are verified.
An initial nucleus of eight fuel assemblies, the first of which contains an activated neutron source, is the minimum source-fuel nucleus which permits subsequent meaningful inverse count-rate monitoring.
This initial nucleus is determined by calculation and previous experience to be markedly subcritical l(k,77<0.95)undertherequiredconditionsoffuelloading.
Each subsequent fuel addition is accompanied by detailed neutron count rate monitoring to determine that the just-loaded fuel assembly does not excessively increase the count rate and that the extrapolated inverse neutron count rate ratio is not decreasing for unexplained reasons.
Criteria for safe fuel loading require that loading operations cease immediately if:
(a) An unanticipated increase in the neutron count rate by a factor of two occurs on all responding instrumentation channels during any single load-ing step after the initial nucleus of eight fuel assemblies is loaded (ex-cluding anticipated changes due to detector and/or source movement), or (b) The neutron count rate on any individual instrumentation channel increases by a factor of five during any single loading step after the initial nucle-us of eight fuel assemblies is loaded (excluding anticipated changes due to detector and/or source movements).
An alarm in the Containment and control room is coupled to the source range channels with a setpoint at approximately five times the curreat count rate.
This alarm automatically alerts personnel of a high count rate and requires an immediate stop of fuel loading operations until the situation is evaluated.
Following completion of fuel loading each assembly and its inserted component will be visually checked for proper location and orientation. TXis JI wel bdude w ve<.h.d 6 sC Rcc A I oc.~+; ms, by gecA rD n ube.rs,
14.2-10 h
ATTACILEYt 2 Form 34634(R845)
DUKE POWER COMPANY NUCLEAR SAFETY EVALUATION CHECKLIST (1) STATION:
h/
UNIT: 1 2
N 3
OTHER:
(2) EVALUATION APPUCABLE TO (DESCRIPTION AND NUMBER OF NSM, PROCEDURE PROCEDURE CHANGE.
OR TEST l EXPERIMENT):
PT/.2 /A-/ 4GSolo.2 C, co re Ved-fica 4 ion
- Procedure c4mye ## O/.
(3) SAFETY EVALUATION - PART A The item to whch this evaluaton is aJ-$ste represent:
%Yes O No A change to the station or procedures as desenbod in the FSAR: or a test or experiment not de-scnbod in the FSAR? Affected FSAR Secton(s) are:
/4 2, /d- /
If the answer to the above is "Yes' identify the affected section(s) of the FSAR. Attach additonal sheets as necessary.
(4) SAFETY EVALUATION - PART B O Yes RNo win this item require a change to the staten Technmal Specificatons? Affected Tech. Specs. Sec-ton (s)are:
N/A If the answer to the abow is "Yes' identify the specification (s) effected and/or attach the appilcable page(s) with the j
change (s) indcated. Tech. Spec. changes requwe NSRB and NRC approval pnor to use, (5) SAFETY EVALUATION - PART C As a roeult of the item to whch this evaluaton is apFaNa O Yes JB1No Wil the probablity of an amriant provously evaluated in the FSAR be increased? Explain: //o' FS M aostilx f m O m S ss / ECGS axo sk nAw Mo $-lwn
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V Oyes %No win the consequences of an acodont provossly evaluated in the FSAR be increased? Explain: _
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Form 34634 p.s 45)
DUKE POWER COMPANY NUCLEAR SAFETY EVALUATION CHECKLIST O Yes % No May the possblity of an accidgnt which,is different than any already evalupted in the FSAR be cre-ated? Explain:
J/ 4 > KCCA 2 m MX i rlwa % - a.a o'u' 5 h tL Yb c<nn dea lidu k rA & t a se w t r~
cum.
" ' ' ' 5e O Yes plNo Will the probability of a malfunction of equipment important to fety prtmou evaluated in the FSAR beincreased? Explain: % 8 CC A-I dd h
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u O Yes JELNo Will the consequences of a malfuncton of equipment important to safety primously evaluated in the FSAR beincreased? Explain:
&c e/
Oyes JELNo May the posshirty of malfunctions of equipment 6Tpcnient to safety different than any already evolu-atedin the FSAR be created? Explain:
h =4m#4 O Yes jiiiNo Will the margin of safety as defined in the bases to any Technical Specificanon be ruiuced?
Explain:
71A E CCd-M b Me z,J etif- /M M (m nizl}L accA-_
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Justification for the answers above (Yes or No) must be prtmded in the above spaces (attach addrtional sheets as necessary).
l An unrumewed safety question le involved if arr/ answer to Part C atxw is "Yes" and NRC authorization is requwed l
3,/2 !PS (6) Prepared by:
Date:
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3/2./4f Dn << I
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