ML070590683
| ML070590683 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 12/04/1986 |
| From: | GPU Nuclear Corp |
| To: | Office of Nuclear Reactor Regulation, NRC/OGC, NRC Region 1 |
| References | |
| 328227-001, DRF-44201, Rev 1 | |
| Download: ML070590683 (22) | |
Text
10/25/06 18:53:48 Technical Functlons Sefety/Envlronrnental Determination and 50.59 Review (Modification, procedure, test, experiment, or document)
- 1. Is this acti&yMocument listed in Section I or II of the matrices in Corporate P;ocedure 1000-AOM-1291.01? .t+. lgves -UNO: .. . -
If the answer to question 1 is no stop here. (Section 1V activitiesldocuments should be reviewed on a caseby-case basis to determine if this procedure is applicable.) This pro- ,
cedure is not applicable and no docurnentation is required. I f the answer is yes proceed to question 2.
- 2. Is this a substantive revision to the activity/document? Dyes i$No (See Exhibit 3, paragraph 3, this procedure for examples of Ron-substantive changes}
If the answer to question 2 is no stop here. This procedure is not applicable and no documentation is required. If the answer is yes proceed to answer all remaining questions.
These answers become the SafetylEnvironmental Determinatlon and 50.59 Review.
- 3. 006s this activity/document have the potential to adversely affect nuclear safety or safe plant operations? qYes GNo
- 4. Does the activityldocument require revision of the systemkomponent description in the FSAR or otherwise require revision of the Technical Specification8 or any other Licensing Basis Document? ayes 5p0
- 5. Does the activity/document require revision of any procedural or operating description in the F$AR or otherwise require revision of the Technical Specifications or any other licensing Basis Document? nyes wo
- 6. Are tests or experiments conducted which are not described in the FSAR, the Technical Specifications or any other Licensing Basis Document? Dyes Qf40
- 7. Does this document involve any potential Non-Nuclear environmental impact? OYes VNo If any of the answers to questions 3,4, 5, or 6 are yes, proceed to EXHIBIT 6 and prepare a written safety evaluation. I f the answers to 3, 4, 5, or 6 am no, this precludes the occurrence of an Unreviewed Safety Uuestion or Technical Specifications change. If the answer to ques-tion 7 is yes, either redesign or provide supportlng documentation which will permit EIF vironmental Licensing to determine if an adverse environmental impact exists and if raaulatory approval is required (Ref. LP-010). If in doubt, consult the Radiological and En-vironmental Controls Division or Environmental Licensing for assistance in completing the evaluation.
N55047 (09-88)
I0/25/06 18:53:48 Technical Functions Safety Evaluation UNIT OrNCS PAGE2OF l1 SE No. 328227-001 Rev, No.- ;+ A - 5 ACTIVITY/OOCUMENT TITLE Drvwell Core Borins .G Repair Document No.
(if applicable)
Type of Activity/Document SantDl&@-a& R e ~acement l in IUnd
-(Modification, procedure, test, experiment, or document)
This Safety Evaluation provides the basis for determining whether this activi~~cument,invOIvesI an Unreviewed Safety Question or impacts on nuclear safety.
Answer the following questions and provide reason@)for each answer per Exhibit 7. A simple statement of conclusion in itself Is not sufficient. The scope and depth of each reason should be commensurate with the safety significance and complexity of the proposed change.
- 1. Is the margin of safety as defined in Licensing Basis Documents other than the Technical Specifications reduced? QYes CjdNo
- 2. Will implementation of the activityldocurnent adversely affect nuclear safety or safe plant operations? UYes !2$Ncl The following questions comprise the 50.59 considerations and evaluation to determine if an Unreviewed Safety Question exists:
- 3. la the probability of occurrence or the consequences of an accident or malfunction of equip ment important to safety previously evaluated in the Safety Analysis Report increased?
ayes Wo
- 4. Is the possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report created? OYes Epo
- 5. Is the margin of safety as defined in the basis for any Tmhnicai Specification reduced?
OYes #No If any answer above is yes an impact on nuclear safety or an Unreviewed Safety Question exists. If an adverse Impact on nuclear safety exists revise or redesign. If an unreviewed safety question with no adverse impact on nuciear safety exists forward to Licensing with any additional documentation to support a request for NRC approval prior to implementing approval.
- 6. Specify whether or not any of the following are required, and if yeS indicate haw It was resolved Yea TRmWWOther No a Ooss the activityldocument require ati update of the FSAR?
x Explain: Commitments are not violated
- b. Does the adlvityldocument require a Technical Specification Amendment? %
Exptain: Commitments are not violated
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10/25/06 18:53:48 Yes TRflFWRlOther No
- c. Does the activityldocument require Y a Quality Classification List (QCL) Amendment?
Explain:
- d. Does the actlvity/document require a review of a commitment as outlined in the site commitment document?
Explain:
. *f. -
- e. Other: N/A (If none, use NA)
This form with the reasons for the answers, together with all applicable continuation sheets con-stitutes a wrltten Safety Evaluatlon.
t Effective Pages
~ i s of Page No. -
Rev. No. Page No. -Rev. No. Page No. -
Rev. NO.
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. . 10/25/06 18:53:48 SAFETY NMUA'ISOW Nuclear Station: Oyster Creek (Cominurtion S W )
TITLE: Drywell Core Boring & Repair 1.0 PURPOSE The purpose of t b i e Safety Evaluation i s t o assess the coneequencea of core d r i l l i n g the drywell s t e e l pressure v e e s e l in order t o be able t o verify the accuracy of UT measurements of p l a t e thickaeas and to eramine the outet.
surface of the drywell wall f o r causes of the apparent thinning indicated by UT. Corrosion on t h e o u t e r surface ie one pO86ibh cause. Sand removed from behind the shell wall will be t e a t e d for evidence o f contamination.
I Thla S a f e t y Evaluation addresses the core boring, the r e p a i r of the core bored areas, t h e excavation and replacement of a portion of concrete only. Two inch diameter cores will be d r i l l e d in t h e v i c i n i t y of the Downcomers and a t '
e l e v a t i o n s 51'+3'and 85'+3' in the vessel shell. The evaluation of t h e ,
problems r e l a t z d t o opergting the plant with the p o t e n t i a l l y thinued drywell s h e l l w i l l be addressed in a different Safety Evaluation.
In a d d i t i o n t o core drilling in the a r e a of the base sand entrenchment and elevations 51'+3' and 85'+3' where the inside face of the shell is accessible, the drywell coGctete floor will be excavated in two l o c a t i o n s to p e r m i t UT measurements of p l a t e thicknese of the pressure boundary wall under t h e concrete floor. Core samples below the basement f l o o r and within. t h e sand cushion will be taken if UT readings warrant, The concrete floor will be excavated i n two areas next t o vent pipes 5B and 11A ( o r 17D). The excavation w i l l be one f o o t wide and will remove the coucrete from the curb edge t o 6 fee&from the pedestal and down t o the drywell shell fu a v e r t i c a l direction.
In order t o monitor the d r y w e l l wall thlckneaa during 12R, an approx.
length of t h e concrete curb and floor under t h e curb ( t o one foot d e p t Y i p t l the v i c i d t y of vent bay 17 will be removed The core bored a r e a s will be repaired and pressure t e a t e d In accordance w i t h the requirements of the Tech. Spec, ASME Boiler and Preseure Vessel Codes Section VI11 1962, Code Caees No, 1270N5, 1271H, 1272N5, o r Section If1 Division I , Sybjectlon NE 1986, where appropriate end 10CFR50 Appendix J.
Section VXkI Division 1, 1986 of the Code w a s used far the r e p a i r , as permitted. The excavation in the concrete floor will be f i l l e d with Dow I C o m i n g S i l i c o n e RTV Foam, topped by a layer of Light Density Silicone 1 Elastomer. The removed portion of the concrete curb a t vent bay 1 7 w i l l not be replaced. The space l e f t by the removal of the concrete under t h e curb w i l l be f l l l c d a8 above.
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2.0 SYSTEHS AFFECTED
10/25/06 18:53:48 SE NO. 328227-001 Rev. 4 Page 4 of 1 1 2.2 System No. 212, Core Spray and Automatic Depressurization System
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2.3 Drawings showing this area o f the shell and concrete:
2.3.1 Chicago Bridge and Iron Co. Contract Drawings 9-0971.
2.3-2 Burns and Roe Contract Drawings 2299 Drawings No. 4059, Sheets 1 and 2 of 3 and #4070.
2.4 Documents that Describe the Drywell ' Structure are 1 I sted below.
2.4.1 Amendment #I5 t o O.C. FDSAR, Primary containment Design Report.
2.4.2 FSAR Updates, Paragraph 3.8.2.
, 2.4.3 OC Technical Specification Section 3.5, 4.5 and 5.2.
2.4.4 10CFR 50 Appendlx J.
2.4.5 CB&E Stress Report, "Structural Design of the Pressure Suppression Containment Vessels" for JCPL/Burns & Roe, Inc.,
CB&I Company Contract No. 9-0971, 1965.
2.4.6 ASME Boiler and Pressure Vessel Code,Section VIII, -.1962, as well as 1986.
2.4.7 ASME Boiler and Pressure Vessel Code, Section 111, D i v i s i o n I, Subsection ME, 1986.
2.4.8 National Boiler Inspection Code, 1983.
3.0 EFFECTS ON SAFETY 3.1 Documents that define the safety function o f the Drywell are 2.4.1 &
2.4.2 as above.
3.2 Safety Function o f Drywell Containment Structure The Drywell Containment Structure (DCS) of the O.C. Nuclear Power Plant houses the reactor vessel, reactor coolant circulating loops, a portion of the main steam line, and other components assoclated with the reactor system. It i s a combination sphere, right cylinder and ellipsoidal dome that resembles an inverted light bulb. The pressure vessel is constructed o f welded carbon steel SA 212 G R . 8 as defined by ASME-1965, having different thicknesses i n different areas. Ten vent pipes, 6 ft. in. diameter and spaced equally along the circumference of the bottom spherical dome connect the drywell and the vent headers i n the pressure absorption chamber which i s called the torus.
32 13H/0090H
10/25/06 18:53:48 SE NO. 328227-001 Rev. 4 Page 5 of 1 1 The primary safety function o f DCS i s containment o f radioactive and contaminated substances that may form during and after the postulated design basis accident ( D B A ) . I n addition, the DCS physically supports safety related systems and components inside the drywell. The DCS should be capable of maintaining structural integrity for a combination of loads a t different plant operating modes, as described in FSAR.
3.2-1 Effect o f this activity on safety:
Technical Specification 3.5.A.2 specifies the conditions which must exist fw- maintenance and repair of the containment. These conditions are:
o Reactor mode switch must be locked in shutdown or refuel o An operable flow path from Condensate Storage Tank (CSTI to vessel o F i r e protection system m u s t be operable o RCS is less than 212°F and vented o At least one core spray pump and associated valves are operable from control room o Torus mechanically intact o No work shall be performed which could d r a i n vessel below 4 ' - 8" above T A f o The CST has a level greater than 30 ft.
These conditions w i l l be satisfied durlng this activity.
There does not appear t o be any formal requirement for a recirculation flow path capable o f returning RCS leakage to the torus. It i s believed that the absence of this requirement results from an inherent assumptlon that such a flow path would always exist. The requirement for a "Mechanically intact" torus in addition to the required CST volume suggests this is the case. The proposed sample removal process would invalidate this assumption. Therefore, although the likelihood of needing a recirculation path Is extremeTy low, it would be prudent t o minlmize the number of openings below the downcomers at any one time and to have available a means for sealing these openlngs in short order, if required, based on any observed leakage from the RCS.
3.2.2 In order t o maintain primary containment integrity during t h i s work, any hole that is not being used will be temporarily plugged before it i s permanently repaired.
32 13H/0090H
, . . 10/25/06 18:53:48 SE NO. 328227-001 Rev. 4 Page 6 of 1 1 For each opened hole there must be an assigned individual to ensure that the hole is plugged with a temporary plug by that person whenever the hole i s not in use for inspection activity or before leaving the work area.
The temporary plug must be able to withstand the hydrostatic presslire and temperature expected fol lowing a loss of primary coolant under the current shutdown condition.In keeping wcith the design b a s i s of the drywell, loadings due to DBA will not be considered rapid .
Spacing o f the replacements-in-kind will be as per Code requirements. Rep1acen:lnts-in-kind will be permanently instal led per Code requirements.
3.2.3 A plug that i s 1.18" high will be used for uniformity. In, the cylindrical portion of the drywell at elevation 85'23'
' the vessel wall i s 0.640"minimum which could result in the plug extending 0.540" beyond the back surface of the wall.
Calculation (C-1302-243-5310-035) shows that no contact with the concrete will occur in the worst postulated design accident, Including jet impingement, provided that the insulation i s removed in an equivalent amount. For additional safety, a l l the insulation at the back of the plugs at elevation 8 5 ' ~ 3 ' wi 11 be removed.
3.2.4 A similar analysis for the spherical portion o f the drywell at elevation 51'+3' was performed and shown to be adequate.
3.3 The core boring o f the presxui-e vessel and concrete floor excavation Ni 1 I not affect the future performance o f t h e drywe1 1 presx1ir.e ve.iFel, due t o the fol ltjwing:
3.3.1 Sys tern Performance:
- 1. The core drilling and subsequent repair will be performed during cold shutdown of the plant. During this time, primary containment integrity is not requlred as per Tech Specs Section 3 . 5 . A as long as secondary containment integrity i s maintained.
- 2. Care will be taken not t o c u t or damage any burried pipe or cable during excavation or core drilling.
- 3. The cores will be 2 " in diameter. Holes will not be drilled near welds. Holes will be spaced such that ASME Section V I 1 1 requirements are not violated. Holes w i l l not be drilled i n reinforcing pads.
- 4. The hole produced by core boring will be permanentlj repaired as per Code requirements. Local pressure testing will be perfoi-med as required by Tech Specs Section 4.5.
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10/25/06 1 8 : 5 3 : 4 8 SE NO. 328227-001 Rev. 4 Page 7 of 1 1
- 5. The sand removed from behind the pressure vessel wall will enable visual examination using fiber optics and will be replaced prior to permanently repalring the openings. I
- 6. The excavation in the concrete floor $ J i l l be filled H i t h ,
Dow Corning 3-6548 Silicone RTV Foam which i s a f i r e 0 barrier material approved f o r nuclear a p p l ications. The top one inch to I 1 / 2 inch of the f i l l will be Light Density Si1 icone Elastomer by Promatec Inc. which provides an effective barrier against water+, gas pressure and fire. In order to prevent water penetration, the I .
vertical and hmizontal joints bethieen the new {.ut i n the concrete curb (bay 1 7 ) and the drywell vessel will be sealed using S i 1 itone Elastomer. I I
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4
- 7. This repair and pressure testing will be accomplished prior to restart of the plant. After the repair, the performance of the drywell pressure vessel w i 1 1 be restored t o that prior t o core boring.
- 8. The integrity of the concrete curb to be cut in bay 17 has no bearing on the stresses i n the drywell pressure vessel.
3.3.2 Quality Standards T h i s activity does not affect the quality standavds of t h e plant since the integrity o f the drywell z1;ell i : d'i;ui'e;l after the repair.
3.3.3 Natural Phenomena Protection Since the d r y w e l l shell i; protected from outside elements by a safety class structure capable of withstanding tornado and hurricane, and since the plant elevation prevents natural flooding these loadings do not contribute to the concerns posed by the activity. This activity does not affect the integrity o f the drywell shell after a seismic event, due t o the repair accomplished. Drilling will not significantly reduce the load bearing capacity of the containment pressure boundary. In partfcular, this activity will not affect structural integrity durlng a seismic event because of the remaining strength o f the drywell. Original strength i s restored by the permanent repair.
3 . 3 . 4 F i r e Protect ion No effects of the core boring o f the drywell shell In the affected area can be found on the fire protection program of the plant.
321 3Hf0090H
SE NO. 328227-001 Rev. 4 Page 8 o f 1 1 3.3.5 Environmental Qualifications No effects are found by the activity since the structural integrity and stability requirements for the plant are assured.
3.3.6 Missile Protection The affected a r e a i; quite isolated such that m i s s i l e protection i s not necessary.
3.3.7 High Energy tine Break; Internal Flooding The core boring will not a f f e c t t h e capacity of the drywell shell to withstand the loads due to these events, because these areas will be repaired and pressure tested to code.
The effect o f the j e t forces on the p l u g s , as required by the original d e s i g n , have been analyzed and found t o be well w i t h i n the allowables stresses f o r both plug and weld (Calculation No. C-1302-243-5310-03O).
3.3.8 Electrical Separation The core boring does not create any e l e c t r i c a l l y related concerns -
3 . 3 . 3 Electrical Is01 a tion
?IC. effs-l:cs tjet-i<l.t>e the core boring does not generate any e I J c t I i c 3 i 1 y r e 1 a red concerns .
3 . 3 . 1 ; ; E l e c t r i c a l l o a d i n g Impact on Emergency Diesel Generators and c: ]+et y Fiu?.e; No effects per above explanation.
3.3.11 Single Failure Criteria No effects on single failure criteria since the structural integrity o f the drywell shell is assured.
3.3.12 Separation Criteria No effects on this criteria due to same e x p l a n a t i o n i n Section 3.3.11.
3.3.13 Containment Isolation Containment isolation i s the primary safety function o f the drywell. A f t e r accomplishing t h e core boring, the repair-and t h e pressure testing, the containment isolation capability of the drywell will not be changed.
Additionally, IOCFRSO Appendix 3 testing w i 1 1 be performed as required by OCNGS Tech Specs Sectlon 4 . 5 .
321 3H/0090H
. . 10/25/06 18:53:48 SE NO. 328227-001 Rev. 4 Page 9 o f 1 1 I
3.3.14 Mater ia1 s Compat ib i 1 i ty This is not a concern since the shell will be repaired using material compatible with the base metal regarding strength, I chemical composition, expansion, and h e a t treatment (identical material is no longer available). The p l u g s used I for repairing the core tiores w i l l be designed, f a b r i c a t e d ,
tested and installed to meet the original ASME Boiler 8 Pressure Vessel Code S e c t i o n VI11 1962 and Section V I 1 1 1386 as applicable.
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Licensing--Basis Oocuments Margin of Safety 3.4 ..--.-.-
The margin o f safety of the drywell shell will not be changed b y ,
this core boring, after the repair and the pressure testing w i l l be accomplished. . I I
3.5 Nuclear Safety/Safe Plant Operation -- ,
Since the structural integrity of the drywell will not be affected by the core boring o f the shell, nuclear safety and safe plant operation will not be affected.
3.6 Probability GfOccurrence or Consequences o f an Accident The core boring o f the shell will not cause an accident and therefore will not affect the probability of occurrence of any accident. Furthermore, ;ince the containment isolation function o f '
the drywell is intact, the consequence Q F an!y portulcited Accident w--.--_i l l not increase.
3.7 -.Pr ob a b - i 1 i ty of Oc c u rre n c e ,, ;x-ccn s e _ p - ~<.of c f u n ct.-i.en ..;f .2 a fe-t y Egu ivent_
Due to the fact that the structural integrity o f the d i j w e l l h d not ~
been affected by this activity, the probability o f occurrence or consequence of a malfunction of safety equipment i n the plant w i l l not increase.
3.8 Possibility for an Accident or Malfunction o f a Different Type Than Previously Identified in FDSAR This activity is a departure from the conditions delineated in FDSAR in that containment is breached by other than a closure device, with fuel in the vessel. However, this condition will not cause more severe accident than those analyzed previously in FDSAR because this activity w i l l not affect the performance o f other s a f e t y related components, systems or structures in the plant, and any drilled hole t h a t is not being used will be temporarily plugged.
SE N.o. 328227-001 Rev, 4 Page 11 o f 1 1 I
4.2 Potential Environmental Impact Since the activity does not affect the environment, i t does not have any potential impact to the following:
A. Environmental Technical Specification 0
- 6. Applicable Environmental Permit Requirements C. Final Envi ronmen'tal Statement ,
D. Env :I ranmental Impact Statement Consequently, no evaluation is required to ;ee i f the above documents are violated. I
. I I 5 .O CONCLUSION In order t o investigate the extent of the corrosion o f the drywell s h e l l at its outside face, to verify t h e accuracy o f the Ultrasonic measurements and to chemically analyze the sand behind the shell t o determine the cause o f potential corrosion, two inch diameter cores w i l l be drilled in the drywell s h e l l . In addition, the concrete floor will be excavated i n two areas for- additional ultrasonic testing o f the shell.
The core bored holes w i l l be repaired by installing welded plugs. Each repaired area will be pressure tested. The excavated concrete w i l l be replaced by Silicone R T V Foam topped by a layer o f Light Density Silicone Elastomer.
A t the conclusion of this w l k :
( I ) The structural integrity of the drywel shell Kill .. not
- be. - affected.
(21 The con t a i nmeri t i so1 a t im ;aFe t y f i i r i c t still be intact. Uon:equently, "Ken2 concerns exist due to t h i s activity.
( 3 ) FDSAR and Technical Specification Commitments have not been violated.
(4) Plant Procedures and Safe Practices are not affected.
32 1 3H / 0090H
DOCUMENT NO.
TITLE D r y w e l l Core Boring and Repair REV
SUMMARY
OF CHANGE APPROVAL DATE Revise to d e l e t e replacement of excavated concrete and replace by.Silicone RTV Foam and L i g h t 1 Density Silicone Elastomer
'In a d d i t i o n , tne effect on the plugs o f t h e j e t f o r c e s due to h i g h energy line break is addressed
.r . ' - .. -,-.
Referenc-e ASME Code Section 111, Division 1, S&bsection NE 1986.
2 Revised to incorporate core boring and,repair at elevation 85'23'.
3 Revised to incorpoxate core b o r i n g and repair at
?levation 51 '53'
,+. 18-2 c
4 ;evised t o i n c o r orate the removal of Tortion of bhe concrete turf: e l e v . 12'.
R c
A 0 0 0 0 0 3 6 12.83
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10/25/06 18:53:48 DQCUMENT NO.
mNuclear 33-328-227-007 TITLE
- Drywell Core Boring and Repair REV
SUMMARY
OF CHANGE APeROVAL DATE 5 Revfsed to i n c o r p o r a t e t h e removal of the concret floor under the removed portion of t h e curb.
r.
mNuclear Document Release Form (Referto EMP-008)
Page of Release Action To: AcAuccf o ReviewiComment wconstruction o As-Builts Procurement Record o OperationsjMaintenance D Hold Construction I
E.d f 3 C%dW/t Home Base 5-3 J 0 Tel 7zve Originator Unit a2 Budget Activity # ,% *tt7 woiso ##
I I I List of Released Items (attached)
Special Instructions References Approved: E 3 p d , z A fd Date: 1 l
cc Original (yellow) to ED & CC A 0 0 0 1 0 8 0 A 5-82
- Nuclear Construction Release Checklist DRF#
WA#
Gt3&&
WO/SO#
I-
-No.
I GENERAL
- 1. Will Construction recognize the end use of the document? Ifno, add special instruc-tions on release form-
- 2. Have associated purchase requisitions been issued? (TAP-011)
Has an Installation Specification been issued? (EP-020)
- 3. 1
- (If yes, DRF #
Has an SDD Division II been completed? (EP-005)
- 4. (If yes, DRF) -#
- Are previous1 released documents affected by this DRF?
- 5. (if yes, DRF1-l
- 6. Have construction funds been approved? (EMP-003) 03 f;o
- 7. Has Human Engineering's review and concurrence b
- ~-
Has Fire Hazard Analysis Input and Status form been completed?
a (If yes, DRF # . , , I (EP-013)
- Has the Nuclear Safety/Environmental Impact Evaluation Summary Sheetbeen releas-
- 9. ed for this mod? (If yes, DRF)-# (EP-016) D,,& sa.
&L~~ETG,~ &.&
If this work impacts the loading of IE equipment (e.g., diesel generator or battery 12.3 G io. loads) has the Electrical Power Manager been advised?
- ~~
IMPORTANT TO SAFETY 11.
- ~~
Are documentls) to be released ImDortant to Safetv? (EP-011. ES-011 and EP-0161
~ ~
- 12. Do document(s) to be released have applicable QA concurrence, if required?
- Is environmental qualification of Class 1E components required? (If yes, System Com-19 ponent Evaluation Work Sheet) -# (EP-031)
- HGEngineering Mechanics reviewed and concurred withany changes to ITS piping 14.
- and/or pipe supports? (If yes, GPUN Memo ) - yC
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STATION TIE-IN
- 15. Have documents been verified? (EP-0091 (If yes, Verification #A If a Fire Hzkards Analysis Report amendment is required, has it been requested? (If 16.
If a FSAA amendment is required, has it been requested?-(lf yes, TR # 1
- (EP-016)
If a Unit Technical Specification amendment is required, has it been requested?-@
18.
- ves. TR U ) (EP-016, LP-004)
If an unreviewed safe$ question is involved, has supporting documentation been pro-19.
- vided to Licensina? (EP-016, LP-004)
- 20. Are special instructions for training included, if required?
If Limited Service Life components or subcomponents are involved, have they been
- 21. identified to Preventive Maintenance?
If components or subcomponents requiring spare parts have either been added or
- 22. removed by this release, has Materials Control, Reading been informed of the change?
- 23. Are plant computer configuration changes required? (If yes, TR # 1
< Comments: (use reverse side of this form) i A0001086 12-85
Work Request Maintenance & Construction Date:.ir/2-al_as, Budget Activiw I) 313 227 Authorized Funding $
Schedule thmmitrnent 0 NRC; 0 Other Authorized work
Description:
U Approval:
Copy Distribution White-M&C Worklofd and Control. Yellow-M&C Document Conrrol. Pink-Originator A0001205.2-82
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