ML20133K936
| ML20133K936 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 08/06/1985 |
| From: | Watson R CAROLINA POWER & LIGHT CO. |
| To: | Grimes B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| References | |
| HO-853222-(E), NUDOCS 8508120426 | |
| Download: ML20133K936 (35) | |
Text
D Cp&L Carolina Power & Light Company HARRIS NUCLEAR PROJECT P. O. BOX 165 NEW HILL, NC 27562.
HXDE-XXX-XXX-XXX HO-853222 (E)
August 6, 1985 Mr. Brian K. Grimes, Director Division of Quality Assurance, Vendor and Technical Training Center Program Office of Inspection and Enforcement U. S. Nuclear Regulatory Commission Washington, DC 20555 INTEGRATED DESIGN INSPECTION 50-400/84-48
Dear Mr. Grimes:
At the conclusion of the IDI team's close-out inspection on July 24, 1985, the team requested additional information for examination. The requested information is submitted as enclosures to this letter and is briefly described below: - Supplemental information for our response to item D5.7-2. - Approved FSAR change regarding setpoint documentation for item U6.5-1. - Approved FSAR change regarding cable tray overfill for item D2.4-2. - Approved FSAR change regarding cesium source term for item D2.5-3. - Approved FSAR change regarding Beta credit for washout for items D2.5-5 and D2.5-6. - Summary of activities by Ebasco Applied Physics (items D2.5-1, D2.5-3 through D2.5-7). - Results of QA audit of Ebasco Applied Physics. - Lists several items which were closed in the close-out i
I inspection. These related to circuit and relay changes.
It is our understanding that implementation of these changes described in our responses may be verified by NRC Region II.
~
8508120426 850806 PDR ADOCK 05000400 ef, O
Mr. Brian K. Grimes, Director H0-853222 (E)
Upon your consideration of the enclosed information as discussed with you during the close-out inspection exit meeting on July 24, 1985, it is our understanding that all IDI items are closed except D2.3-1, D2.7-1, D2.7-2, and D2.8-1.
One of these items (D2.3-1) relates to the contain-ment building sump for which we are trying to arrange a meeting with NRR on August 15.
We appreciate the team's efforts in working with us during the close-out inspection. We are available to provide any other information necessary to complete the closure of the remaining items.
Very truly yours, R. A. Watson, Vice President Harris Nuclear Project Department RAW /EMH/ jam Enclosures cc:
Mr. B. C. Buckley (NRC)
Mr. G. F. Maxwell (NRC-SHNPP)
Dr. J. Nelson Grace (NRC-RII)
Mr. Joe Joyce (NRC-ICSB)
Mr. Travis Payne (KUDZU)
Mr. Daniel F. Read (CHANGE /ELP)
Wake County Public Library Mr. Wells Eddleman Mr. John D. Runkle Dr. Richard D. Wilson Mr. G. O. Bright (ASLB)
Dr. J. H. Carpenter (ASLB)
Mr. J. L. Kelley (ASLB) 2
ENCLOSURE 1 INSERT #1 The 480V Class 1E bus undervoltage relay settings will also be reviewed.
l INSERT #2 1
The undervoltage relay on safety-related 480-volt bus IB3-SB was set at 80 volts pickup on the secondary side of the 480/120-volt potential t ransfo rme r.
This equates to a minimum bus voltage of 320 volts and is too low (70% of equipment-related voltage) to provide adequate protection for the equipment fed from this bus. This setting does not agree with the results of the calculation performed to determine the setting of the undervoltage alarm relay, which established a minimum voltage setpoint of 100 volts pickup for the undervoltage relay..This calculated value corresponds to a bus voltage of 400 volts (87% of rated). Also, the selection of the minimum time delay setting on the undervoltage relay may result in nuisance alarms during diesel generator loading.
INSERT #3 The undervoltage relays provided on the 480 volt Class 1E power centers are for load shedding Class IE motors (Power Centers 1A2-SA and IB2-SB) and for alarm (Power Centers 1A3-SA and IB3-SB) whenever there is a loss of bus voltage. The undervol tage alarm function is not intended to alert the operator of a degraded voltage condition, but ra ther that i
power is not available to the motor control centers. Degraded voltage protection is provided by undervoltage relays located at the 6.9kV Class 1E Buses 1A-SA and IB-SB and are set in accordance with criteria of Branch Technical Position PSB-1, " Adequacy of Station Electric Distribution System Voltages."
i The criteria for the undervoltage relay settings on the 480 volt Class 1E power centers is as follows. The dropout settings on all undervoltage relays of the 480 volt Class 1E power centers will be set below the dropout settings of the 6.9kV bus primary undervoltage relays as determined by the latest voltage study, " Adequacy of Station Electric Distribution i
System Voltage" (Feb. 1985). The time delay setting of the undervoltage relays of all 480 volt Class IE power center buses will be selected to ensure:
a) avoiding nuisance relay operation for feeder faults by allowing the feeder breaker solid state trip device to trip first and isolate the fault from the power center bus, and b) for LOCA followed by 4
loss of off-site power conditions, the undervoltage relay to initiate trips of the 460 volt motor breakers prior to closing of the diesel generator breaker.
The re fo re, in accordance with the above, the 480 volt undervoltage relays will be set at 70V (58% of power center bus rated voltage of 480 volt) which is below the dropout setting of the 6.9kV Class IE bus primary undervoltage relays as determined by the latest voltage study
" Adequacy of Station Electric Distribution System Voltages". The time dial position will be set at 2.
5.7 MOTOR ELECTRICAL PROTECTION In response to these concerns, the relay calculations for each of the 460 volt Class 1E motors will be updated and verified using motor specific starting times and safe stall times. The relay calculations will be incorporated in a Standard Relay Calculations and Data Sheet as required by the Harris Plant Engineering Section guidelines for relay protection, which is currently being developed. Accordingly, any missing assumptions will be identified INSERT #1
_and verified. [Co cernin Ine 45U voi t Cl s IE bus u derv ltage
}
resay setti g, t late vol go s dy, Adequa o Sta on Elec ic D stri tion stem olta " (F b,198 sd enni d tha the lay dropou sett gs o 100 olt fo tw of e 48 vol f
C1 s1 powe cente wer acce bl and th re yd peut s tin for the re inin two 0y t Clas 1E owe cent s wi re sed o 110 lt.
he d win for t e s jec sett gs 11 e
ised o ref ct t cor et s unde ol age elay tti s and he t e del set ngs th underv e re ys o all lass )
1E ower enter uses ill e re ewed e ure j oidi o nui ance I fu av o eratio unde tran fent ndervo taae cond'tions. 'The Harris
~P' ant Engineering dertion will develop a design guicei m: and procedures for relay protection.
The Startup Organization will revise Procedure 1/2-9000-E-05 to require the bus voltage to be recorded when measuring the motor running current under both the loaded and unloaded conditions.
In addition, the appropriate process will be established to require the startup organization, upon request from the Harris Plant Engineering Section, to perfonn additional tests on large 460 volt and 6.6KV motors for which specific data cannot be obtained from the vendor.
5.8 CABLE DESIGN AND ANALYSIS The IDI team reviewed cable sizes to ensure cables were properly sized for both normal and overload conditions. Power and control cables were " reviewed for both ampacity and voltage drop considerations. The IDI team identified certain cases where cables were not adequately sized for the worst case voltage conditions.
I Deficiency 5.8-1 stated that when the de system is at minimum voltage, the power cable feeding the de motor operated valves is not adequate. Corrective action has been taken by revising the cable i
size.
The basis of Deficiency 5.'8-2 was the possibility of simultaneous operation of multiple relays in the auxiliary relay panels. Under worst case conditions, several relays can operate concurrently, which might result in a high inrush current (approximately 125A) through the cable feeding the panel.
Due to this high inrush current, the voltage at the relay coils may not be adequate to operate the relay. The Ebasco electrical group has Just completed an evaluation of all ac an'd de control loops. The objective of this effort was to ensure that all associated cables were properly sized. As a result of this evaluation, in conjunction with 05.8-2, the subject cable size has been revised.
32
i 05.7-2 (DEFICIENCY) 480V BUS UNDERVOLTAGE ALARM I
DESCRIPTION SERT Accor ng t the ojec load stu of I 2, the.ini m all able teady sta vol ge of he 4 V Class E Powe Center Bus 3-SB uppl ing M j
lo s oni ) was 28V
.e. 90% f mot rated ol ta or 4 V pl s 3%
r f
c le y tage rop o MCC mot s to e PC b s).
he cal ulat bus derv tage elay ettings re: d p out f 10 (corr spon ng t 400V us olta
) an a ti dial po ition
+
How ver t rel settin for 1 480V ower enter PC) uses e
i orrec y en red in e rel sett g dr ings a 80V nd ti di p siti l in tead of e abo e calculated etting.
RESPONSE
There are total of four 480V Class lE Power Center Buses and all are affected.
CINSERT#>3 The latest project voltage study, " Adequacy of Station Electric Distribution 7
System Voltages" for compliance with BTP PSB-1 of February,1985 which was not available at the time of the IDI inspection, revises the minimum allowable steady state voltage for PC buses l A3-SA and 183-SB (both supplying MCC loads only) to 444V and that for PC buses l A2-SA and 182-S8 (supplying motor loads only) to 424V. Considering that the preferred drop out setting of the bus undervoltage relay is just below the bus minimum allowable steady state voltage, bus PT ratio (480/120V), and the relay available taps (60,70,80,100&l10V - not continuously adjustable) the relay dropout setting of f 100V was acceptable. However, in view of the revised values for the minimum allowable steady state voltages for the 480V Class 1E PC buses per the voltage study of 1985, the dropout settings of the undervoltage relays for buses l A3-SA and 183-S8 will be revised to 110Y.
The relay dropout settings of 100V for PC buses l A2-SA and IB3-SB are considered acceptable.
i l
The time delay settings on the undervoltage relays of all Class lE PC buses will be reviewed to e'nsure avoiding nuisance relay operation undir any utransient undervoltage conditions.
The relay setting drawings for all 480V PC buses will be revised to indicate the co'rrect bus undervoltage relay settings.
il The Harris Plant Engineering Section is developing design guidelines for relay protection including that for all 480V PC buses undervoltage relays.
i l
S Y'LAs if b
-1 EJCLOSURE 2 1lt
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FSAR A'iENi/MST REVIEW Al PROVA1. FORW.
(P.AF) g F3 A -D'7 ITEM NO.
373 m
DI dew 5hms F - 30 2.
DuE DATE Re
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FS AR/ER SEC110N TO BE AVENDED:
/,8 k /, /d CPSL REVIEW RESPONSIBILITY /PR'IMARY SIGNATURE REVIEWER:
jf'l N WyJ5M F EC3NENDED CF.ANGE:
( At t act. e* Ne! up FS AF/ER pare's) f or clarity) y A v b.
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t REASON FOR REVISION:
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In i t ia1 Re vi e.ee r s:.
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QDVANCE COPY ID E8ASCO SERVICEJ INCORPORATEa EKO Two World Trade Center, New York, N Y.10048 JUL 2 41985 EB-FC-827 File No.:
ll.Q.D.6 Mr L I Loflin, Manager Engineering - Harris Plant Carolina Power & Light Company P O Box 101 New Hill, North Carolina 27562
Dear Mr Loflin:
Subject:
SHEARON HARRIS NUCLEAR POWER PLANT FSAR CHANGE NOTICE NO. F-302 Attached for your use and approval is FSAR Change Notice No. F-302.
This change notice revises Page 1.8-136 in order to clarify the program in place for determining cafety-related, non-technical specification, BOP Setpoints as per IDI Item U6.5-1.
It is recommended that this change notice be incorporated into the next FSAR Amendment.
Please advise us of your comments and/
or approval of the subject change notice.
In accordance with Ebasco's procedures, this change notice will not be considered in effect until formal CP&L approval is received.
If there are any questions, please advise.
Very truly yours, I
h Urf0lMWfUX&~
DMR/vhr A C Anderson L
Attachment Project Manager cc All with Attachment L I Loflin M Thompson E Harris D McCarthy N J Chiangi J L Willis A T Parker M Shannon (2)
I I
RECEIVED SAR CHANGE REQUEST
, m,f er,
'4 CHMGE NO. # 36rD.
D. E. CONNELLY TO W Malec FR0g W Pehush
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PROJECT LeC EN elN G ENGtN E ER b E AD DisCIPUN E EN GIN E ER CLIENT _ Carolina Power di Light Company PROJECT Shearon Harris NPP REQUESTED CHANGE:
1.8-136 PSAR X
FSAR Section(s) 1.8 Page(s)
Recommended change and reasons: (Note - Attach madedsp copy of affected SAR pages)
Clarification of actual program in place for determining safety non-tech spec. BOP setpoints. This is a follow up to the IDI Unresolved Item U6.5-1 which is now closed.
m, j
Approval Date
(
/ - 2 Y',-g [
Approval Date ysciPuME suPERwisoN LICEN$lMG REcomENDATI Submit with next SAR amendment Reject proposed cha1ge Hold for FSAR preparation Other f:.
[,No Notify NRC Yes Comments Approvat _
g4 Date
[JM *h '
DISPO$lTION:
Letter to client /88#<*6Z7 Client lett r Implement Do not implement Comments
(
cc: Project Enginwr TO BE RETAINED IN PLE FILEl I
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SMNPF FSAR a
4 1
d)
The accuracy of all setpoints is equal to or better than the accuracy assuaed in the safety analysis.
Instrisnent intervals are chosen for tii.
desi n conditions in which they are installed in order not to anneal, stress d
relieve, or wrk harden to the extent that they will not maintain the required accuracy.
Design verification is included as part of the equipuent qualification program as recoismended in Regulatory Guide 1.89.
e)
Instrtsaents important to safety have securing devices on the setpoint adjustinent mechanists and/or are under administrative control. The securing device is designed such that during securing or releasing it will not alter the setpoint.
f)
Voctanentation of assumptions used in selecting setpoint values and minimura margins, drif t rates and test intervals is contained in the Technical Spec i fica tions. Chapter 15 also contains docunentation of asstusptions used in nelecting setpoint values.
g)
Safety related setpoints not covered by technical specification have sufficient documentation to support the setpoint value, tolerance, and margin to system process limits.
i i
I 1.8-136 I
I l;
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ENCLOSURE 3
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FSAR AMENDMENT REVIEW APPROV/ L FORM (RAF) 4, ITEM NO.
MPE,5 -400
[5 DUE DATE Sl2 /gg
SUBJECT:
psAR c,8W6 Na tC.E F-3d FSAR/t41 SECTION TO BE AMENDED:
9.5 I CP6L REVIEW RESPONSIBILITY / PRIMARY SIGNATURE REVIEWER:
M. F. THornP5cd RECCT! MENDED CHANGE:
( At tach marked-up FSAR/ER page(s) f or clarity)
ATTetdeb
- l s REASON FOR REVISION
T DI.
lTC-M T 2*4 O Initial Reviewers:
U-h-87 m!)f,. sh,/K Initial Reviewers:
it,
J Primari Sig5a'ture Revied /Daff M.f
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(8 600PR0/c f r)
' a u..n SAR CHANGE REQUriST g<
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CLIENT $bbDLINA TQ.WW -h l$NI[0. FRO)ECT SllEA/ON NAA$l S REQUESTED CHMGE:
sxa ZrsAR secius) t 9.* 5. /
Pqe r_ S. 5. /-50 e
Retommended ctarge and reasons: (hete - A:taa markedsp con of affected SAR pages) _.T..b. I F/mdMe 0
- h. 2. 4 -2 cgsis r/Ay oystpm
. See 1m.rert1,
.A$a.4 eA.
A;;reval _
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Date 7 ~ 3 /- 86 e
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M Date 7~3/~&
Ap;roval _
o LICENSING RECOMENDATIONS:
~
Sutet mth next SAR amend ent Reject proposed cha ge Hold for FSAR preparat on Other _
Notify NRC Yes hc Comments j
a.
Approva' d
Date 6
- I - 8C Di$PO$lTION:
Letter to client E8-R - 6h Client letter ls;tement Do not impfecert Comments cf. Fige:t Engmet-TO BE RET Alh ED lh PLE FILE 1
V :,wt SHNPP FSAR
(
combustible materials which were representative of heat values of specific materials grouped within the class. These are:
Ordinary Cambustibles 8,000 Btu /lb.
Coc:bustible or Flamable 20,000 Btu /lb.(108,000 Liquids BLu/ gal.)
Diesel Fuel Oil (No. 2) 140,000 Btu / gal.
Charcoal 10,000 Btu /lb.
Combustible loading for minor amounts of grease, integral with equipment, not exceeding one Ib. each, was not inventoried since it does not create a significant fire hazard.
Using manufacturer's data on specific cabic construction used in SHNPP and the Btu content of the insulation materials, Btu values were derived for each running foot (RF) of 24 in, wide, 4-inch deep, 40 percent loaded, power, control and instrumentation cable trays.
These are:
Power 200,000 Btu /RF Control 170,000 Btu /RF Instrumentation 155,000 Btu /RF 20 These values are adjusted proportionally for trays of different width or cable tray loading (fill depth). Maximum allowable fill is assumed and is based on plant design criteria of 30 percent for power trays and 60 percent for control and instrumentation trays. The running foot (RF) value reflected in the fire hazard analysis represents cable trays of various widths and depths and should only be used as reference to determine the linear feet of tray.
The combustible loading for all cables routed in conduit, cast concrete trenches, or contained within metallic cabinets or consoles was not inventoried since they do not create a fire hazard.
Based on the above, three fire areas were identified where the Stu content exceeded 240,000 Btu per square foot, assuming all electrical trays filled to maximum capacity.
These fire areas were Cable Spreading Rooms 1A, 1B, and the Auxiliary Control (Panel) Room. In order to verify a more accurate combustible lo'd in these fire areas, as opposed to the critical maximum allowable load the actual cable tray fill as indicated in the Cable and Conduit List i. e used to calculate average actual tray fill for each cable tray within < 4_h of these three fire areas. The percent fill was then used to calculate the combustible load in these three fire areas and represents the actual cable tray fill percentage, including approximately $ percent additional for potential future use. The resultant combustible loads for these three fire areas are shown in Appendix 9.5A.
k fIn addition to the combustibles normally present in an area, transient ec%s t i ble s wr i ch tright realistically be introduced int e areas as a part of plant.ed opera: en are considcred, as detailed in fire hazard analysis for each HN SE AT I QHached; 9,5.1.so a_n, w. yo
F Sc4 INSERT 1 T. 4'eReci 4 k c_
The cable insulation combustible loading will bc (_f5fs4[>
__ -' 7 '--
uLl"i*'tw!r.%, a mm u ma a au a Conduit List will be used to determine actual cable tray fill.
In certain instances, actual cable tray fill may be permitted l
to exceed plant design criteria.
Since this situation is not expected to occur for the entire length of a cabic tray contained in a fire area or zone, the actual average fill for that cable tray is not expected to exceed the values assumed in the com-bustible loading calculations.
If the actual average fill for a cable tray exceeds plant design criteria, the actual average fill for all cable trays in the respective fire area or zone will be calculated.
Should the overall actual average cable tray fill for the fire area or zone exceed plant design criteria, the combustible loading calculation and Fire Hazards Analysis will be revised accordingly.
.. -. ~ _ -. - - - - _ _ _. - _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
ENCLOSURE 4 M
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FSAR AMENDMENT REVIEW APPROVAL FORM (RAF) 1 TEM NO.
N966"40 i DUE DATE gl49 SU BJECT:
F5A9:. CHMGE r4cTsE F-3o3 FSAR/ER SECTION TO BC AMENDED:
t"L. % t, 11' 3.2.
CP&L REVIEW RESPONSIBILITY / PRIMARY SIGNATURE REVIEWER:
m.F. h esod RECOMMENDED CHANGE:
( At tach marked-up FS AR/ER page(s) f or clarity)
ATTACH &O REASON FOR REVISION:
ItI item 9 2.5-3 i.
,v Initial Reviewers:
M'6/J/66 G
/
Initial Reviewers:
E d
Priinfry Signature Reli@er/Dafe M. F. THowPsod, drR.
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(8 600PR0/c f r)
~cv vu m jy( p,4 g 5
'5AR CHANGE REQUEST aneu..n D. E. CONNELLY otmGE n0. / -Et 3 b
W H
FROW T0
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C b 5b-b PRORCT M AON 6 R\\E CLIENT REQUESTED DIMGE:
]FSAR Sectiar(s)01
- 2.* I
- I 2-Page's) O
- l' I~ k PSAR
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fo?. 3, el = l]__
Re:ommended char.ge and reasons: (hete - Atta:t madedsp copy of affected 1AR pages)
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AS dAdEM
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299 E-Soutte SkYe_
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_Reut ce4 7 cme _s
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id1.s T"A. fel.iei edy u;4 tb e45ve n T
P~3e-F:d 7 Sc om n;tevai
\\) _ cult:7 oate 7/ n-/ 65~
g,cva' Date N,
3 UCEh$thG RECOWEMDATIONS:
Scte:t with r. ext $AR mendment Reject preposed change Hold fer FSAR pre;a ation
[
Oeer h:tify NRC Yes No
(
Comments Th ddb r/ 7~tM i2.1. l - 2 E addaar ID r 4A z 3 g U
l
< ep 4 _p_ 9L ~6 d.dMsh4I 4'X l
/*> 1 km A,51r.L._GM "
A;;tova' Date P.osECT L a C E N.iN G E Pe G,h t g.
DISP 051Tl0h; F/'/8 #
Letter to che.! / A- /~/
7.3.S~
Client lettet Implemer.t Do not splemert Cot.? ents
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c: Prye:: E g rto TO BE PET Ath ED IN PLE FILES
SHNPP FSAR 3
ti me. The activity concentrations on both dekineralisere reflect annual resin
(
replacement.
(
12.2.1.11 Reactor Startup Neutron Sources The primary and secondary reactor startup neutron sources (see Section 4.2) remain in the reactor for the duration of their useful life and thus are not considered as a separate radiation ahtelding problem.
Only the primary reactor startup neutron source is initially radioactive and is adequately shielded when shipped to, and handled at the site.
12.2.1.12 Post TMI Shielding Review Accident Sources The ource terms used its the estimation of radiation doses are losed those a
sup;11 by WestinEhouse for this TSAR shown in Table 12.2.1-23 a those required b he h*EC (kegulatory Guides 1.3, 1.4, 1.7, ar.d SRP
.6.5).
From the total cor 4nventory, 100 percent of noble gas
, 50 percent of halogens, and one perce t o.'
the other solids are te determine the source terms for the reactor coo t.
These are tabul acccrding to energy groups at various time intervals in kle 12.2.1-25 n addition, the source terms have also been determined which t into ccount the dilution by refueling 3
water storage tank volume and are sho in Table 12.2.1-26.
The gaseous source terms for g radiation hin the containment atmosphere are based on 100 percent of
( core noble gas in tory and 25 percent of the
(
4 core halogen inventory, iese are tabulated in Table
.2.1-27.
For the long ter adiation sources, 20 percent Cesium icven is added.
These are sho - in Table 12.2.1-28 at time one year after reless f activity.
> 1t ms emphasized tha't ~ 100 percent noble gas core inventory is inclu d in sou terms for liquid systema as well as for containment air. Hence, the culated dose rates are extremely conservative.
0 l
12.2.1-4 3
y.s SHNPP TSAR 3.1 years. Activity decay is taken for the minimum waiting period of three days following shutdown.
c)
The shielding design of the f uci pools is based upon the loadings described in Section 9.1.
Shiciding for the Spent Fuel Fool Cooling and Cleanup System is based upon source terres derived f rom normal system operation as described in Sections 9.1.3 and 11.1 in conjunction with approximately two-thirds of a core i
15 l stored in the pool and fuel clad defects in the fuel rods which generate one per:ent of rated core thermal power. Refer to Sections 9.1.1 and 9.1.2 f or a discussion on the capacity of the fuel pools.
12.3.2.14 control Room For the purpose of designing control rose shielding, the radioactivity releases f rom the maxistar loss-of-coolant accident (LOCA) are controlling.
The two sources considered in designing the shielding were a)
Direct gamma radiation from the containment atmosphere and from radioactivity collected on emergency filters.
71D-14844 (Reference 12.3.2-8) source terms, release f ractions and plate-out fractions are considered. A eeee-eenteinerentdeak Tate-is messaned Jor the duration.of the--accident.--A-uniform distribution 1f -radioactivity within the Containment.1s.assmed. - No credit is taken for iodine removal fros the atmosphera hy sprays.
Credit _Jor. Lost-accident. decay is considered.
- . t
.7,0.5 e b t-
.2 the direct whole body dose to control room personnel following a LOCA is t
computed to be less than 3.0 rem in 90 days. Credit for 8.5 ft. of concrete (4.5 f t. for thecontainment shield wall and 4 f t. for the control room shielding) was taken. A minimum shield thickness of 6 ft. separates the energency (11ters from the Control, Roos.
l' In addition to shielding from external exposures, the Control Roos is designed to operate in an isolated mode under accident conditions in order to minimize the quantity of airborne radioactivity which enters the Control Room and thereby ensure compliance with CDC 19 of 10CFR50 Appendix A.
A detailed description of the system design is provided in Section 9.4.1.
Chapter 15 includes an evaluation of the exposures to control room personnel f ollowing the design basis accident.
b)
Direct gamma radiation f rom radiation IcakJEe external to containment in addititm to direct shine exposure from airborne radioactivity in the Containment folloving a design basis loss-of-coolant accident, control room personne' may rect iva a small crposure due to external exposure to the passage of the t neous p* me which could result f rom containment leakage. As sirsing the Tits.4844 soo ce terms, an integrated containment leak rate of 0.1 percent per day 'or the f trat 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post LOCA and 0.05 percent per day for the renalndre of thc accident and five percentile voeteorology, the integrated esposure (30 days) to control room personnel f rom this source of exposure will not exec 2d 0.1 rem.
(
12.3.2-12 A m&. nt W,. I 's
F-303 INSERT A 12.2.1.12 Post TMI Shielding Review Accident Sources The source terms used in the estimation of radiation doses are based on the Westinghouse Radiation Analysis Manual (WRAM-Refe-
)
rence 12. 2.1-4).
The plant specific core design parameters are stated in WRAM Table 5-1.
Noble gas and halogen inventories are given in WRid Table 5-9; while solid fission products f rom the I
spent f uel fission products inventory at shutdown are given in WRAM Table 5-29.
The values are converted to a total core inven-tory basis.
In addition fission products not mentioned in WRAM but which are described in Reference 12.2.1-5 have been included in the evaluation.
Decay daughter dose contributions for all fission products have been included in the inventory.
The model assumoo removal by plate-out using a mechanistic model considering elementel, particulato and organic fractions of solid fission products.
Plate-out source strengths in containment are given in Table 12.2.1-25.
The model also assumes mechanistic spray removal and dilution of the released 50 percent halogen core inventory and 1 percent solid fission products inventory source terms with the combined volumes of the Reactor coolant, accumulators and the refueling water storage tank.
The resulting liquid acti-vity as a f unction of time is given in Table 12.2.1-26.(Note to CP&L - Thic Tabic was revised via F-299, EB-FC-825).
The gaseous source terms for gamma radiation within the containment atmosphere are based on 100 percent of the core noble gas inventory, 50 percent of the coro halogen inventory, and 1 percent solid fis-i sion products inventory.
The assumptions of removal by plate-out and spray are as stated above.
These values are tabulated in Table 12.2.1-27.
i INST,RT B 12.3.2.14 control Room 1
The model assumos removal by plate-out using a mechanistic model con-l sidering clcmontal, particulate and organic f ractions of haloacnr as J
well as the particulate fractions of solid fission products, Plato-ott source strengths in containment are given in Table 12.2.1-25.
J c edit in taken for iodino removal from the atmosphero by sprays, i
c edit to post-accident decay is considered.
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ENCLOSURE'S
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FSAR AMENDMENT REVIEW APPROVAL FORM (RAF) 1 TEM NO. 4965-40 L DUE DATE SU BJECT: VSRR CMt4GE e4cnce F 2c(1 FSAR/ER SECTION TO BE AMENDED: 3. ll A; 111-CP&L REVIEW RESPONSIBILITY / PRIMARY SIGNATURE REVIEWER:
(Y).F. THemP504 RECUtMENDED CHANGE:
(At tach marked-up FSAR/ER page(s) f or clarity) l PTTACH60 REASON FOR REVISION:
l IDI 17 @
"D F f-5 l
Init tal Reviewers: h Yp/lI 7
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Initial Reviewers:
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ESASCO SERVICES INCORPORATED EBEO TwoWorldTradeCenter.New York NY 10049 g jg g ES-FC-825 File No.
I1.Q.D.6 Mr L I Loflin, Manager Engineering - Harris Plant Carolina Power & Light Company P O Box 101 New Hill, North Carolina 27562
Dear Mr Loflin:
Subject:
SHEARON HARRIS NUCLEAR POWER PLA'T FSAR CHANGE NOTICE NO. 299 Attached for your use and approval is ISAR Change Notict 50. F-299.
This change notice revises pages 3.llA-5, 7 and Tabic 12.2.1-26 as a result of using a mechanistic model for the removal of activity by plate-out and sprays.
It is recommended that this change notice be incorporated into the next FSAR Amendment.
Please advise us of your comments and/or approval of the subject change notice.
In accordance with Ebasco's procedures, this change notice will not be considered in effect until formal CP6L approval is received.
If you have any questions, please advise.
Very truly yours, ho n uG~-.
AC nderson Project Manager DMR:maa Attachment cc:
All with Attachment L I Loflin M Thompson E Harris D McCarthy N J Chiangi J L Willis A T Parker M Shannon (2)
ENO JJL 251995
. a m/i n.
SAR CHANGE REQUEST CHC:GE NO.
T-d//
M TO
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CLIENT C PLL eRatCT _ SHE A rom H A R R) S REQUESTED CH ANCE:
PSAR FSAR Section(s) APPEG1x 3 1) A p g,3) '3. )) A-5,-7 1 2 1.. l l 2..
o Re:ommended cta ge an: reasons: (heti - A:ta:h rearkedsp ccn of a'fected 5A: gages) _9_qc_S PncK.) ' -
teS tesud og_usset w e c k w' M ic. w ocj e.).let h Asct J
V TOWoVCd o O.C$ t'N1's k
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Approva!
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OllC i P.l~ L SiJ P L W t 9 ; et LICENSING RECOWENDATIONS:
V Sukit w.th next SAR amendment Fe,e:t picposed chrfe Hold for FSAR prepa ation C e-Notify NRC Yes YNe
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DISPOSITION:
Letter to client EA 'Al-Ef
Cae t letter inlecent De riot implement Comments ec n:: :t Ernen TO BE RET AlhED IN PLE FILES
SE5?P FSAR APPEcla 3. IIA M.EEG 0585 0388MI53N f
CATEG'F' II Acclicable tc Ecs'paen* Oaall'Ise in f.ccoccance olth IEEE Stc. 325-1971 St e e cr ue -r i s nove l es* Po.e
- Pl e t Pr c,:* ar 1
car neletain its requirac f uncilonel ope'et t lity i t its su f ace tempeesture reaches r
the calculated value or (II) r ec uell f i cat f or. t es t ing be pe* f o esc w it h ac trop
- l a t e margins, or (lit) cuttitled phys i ca l pr et eet t or be prov'ded to ass.ro that the su* f ace teg eratu o will no* exceec tne actual cualif icat ice t enc e*ature.
1.3 Ef f e:+s o' Cne rice' Soray Tne ettocts of caust!c spray should be 1.3 The most seve o contalrunent spray acc esses for the oculpment cue i t t icat t ori.
env ir onment (boron concentration anc pH The concoa.tration of caustics use: t or level) is usec f or environmental cue t if ication sho le be ocul a'est to or quellfleetion. The actual lea l cula+ec )
(
so,e s evere,h.,,ho,e u,ec i n,h.
,, ant s,,a, env,,o _ t b _nc, en,,ostut...c y
coate'. neat spray sys*en.
If tne chotical single tailure, cogosition of the caustic sprey con be affectec pr ocalpeent mo ltunctions, the ecs? severe caustic speay environner.t that results from a single tallure In the spray syster shwld be assumes.. See SalP:Section 6.5.7 (NJRE".r-75/067). peregraph I I, Ites le) for caustic sprey solution guidelines 1.4 Radiettor Conditions inside anc Outside Co*.tal vent The raciation envirorument to*
1.4 For cualifica* 6on purposes, res;ctiors qua lif icat ion of eculpee.t should be bese:
In air dose due to spray washoot one on t ne n o9ma l ly ecoctec regiation plateost arem usec in calculeting environment ove* the oculpment quellflec the post-accident radiatlor, environments, life, plus that associated with the most The ef ore, ractation doses used in seve o des ign bests occident (CEA) during qualif icatior, are maximum tota l int egr atec or f ollarleg which that equipsest amist dose calculated over the equipment remain f unctional, it should be assumed quallflod life, plus that associated with that t he CEi A ref etec ere.Irorimeatal condi-the most severe design basis occideatt.
t
)
tions occur at tse end of two azulpeent cu m isee :Ite.
e 3.11A-5 Amendment No. 16
SES7? UER APPf.eCl* 3.11A ed EEG-058t 03*MAI50*i cat [,3Rv 11 Ac g l i ca t i e te E4.,; pas
- t Cheellfled in Se na e or ka-e l s A c t ee ' P.w o
- P i a-- Pr og e-Ac cor de-ee a l ta s TEE 5+c. 323-1978 the sayee ll.
1he assJaption of the steer generators an the reactor vessel and since the Contelnas-t Spray welf ore distrltm;tior of activity and/or the containment ventilation and t he cts; % o.' t he c omte ; eumo r t a
- t ime filtrattor systoe. Provide e' tac f or zero 16 #wot opprop-late, the coctalrument atmos,* re, a dete mination was esce to esssue a unifore distrilmation of activity theowgwoct the containment.
(a) Crecit f o-the r em:>ve l of s i emorn e (4) Et'ects of E5r syst ems, suen as cor.t a i reme nt s pr ay s a n c cont a i reme n t activity try ESF syste.s has pee-tanee.
In sedition, the dic-Itution weetilation anc f iltrat tor systems, e
- i ct act to rmore eleborno ectivity of activity is taaer into account as en: re:Istritmate activity within desc-lee in (3) adeve en: en ($1 cor.t e t rume n t, shw i e be ce i cu l at ed
- belor, g
using the same assustlor's usec In 0
tne calculat f or of of f site dose.
See 55to Section 15.6.$ (NJs[*.r-75/067) anc the rdlated sections ref erence:
la the Appencices to that section.
wsERT A 16 (5) Be t w a l d epo s t t i on t i.e.. p l ate-ou t )
(5)
^5ww soce s ai s. ries re c re'ev for : st e-out ; ha everi'ine cor i n-o airl>orne eetivitt shouie t.
ment 5 - sauc e. <.rus a-o eiop.:
ee+e in.: using a mechanistic :.aesi se: best es t lee tes f or the su>ce l t>g a s s u't.
dilut on of p**ce*.t of p a
- eme
- ers.
The essen; tion of 50 the core in mtori of
.ogens an:
i ellees alth the po-cent Instantaneous plate-cut of 1 po cent of o cambined volume
- the Reactor the lodlne reteesoc f rcer the core 6%oald not be ende. Removal of Coolaat. Acc-. t ai o Boror injection lod!ne f rom su f aces try stee, 5se ge Ta n and ti e Re elln; eate*
r le concessate flor o+ oeshot t try the St or e ank.
TN resul. ; initial c or.t e i me n t s p* a y me, t>e a s s wee d i f sJe.ci tutec coo eat) a ct l ty is or on FSAR Tat e 12.2.1-26 sect ef f ects can tue justif lec and e d nt i t led try entlysis or o c erleent.
(6) For veistle19ec och gee nt located le (6) The gamut dose and dose rate use: in the cont e t nese t, tt e gamme aos a ont dose evalif ication f or equipes*.T locate:
ra's shoJld be equal to the dose and dose laside contelneont is calculatec f or weelous tones utillaing distance and rate at the cor.te* point of tta con-tainment plus the contritn;tle from shleiding credits. 8tef er to F5AR Appenclu 3.11E for applicacle coses in location deponoemt sou*ces sc h as the sarc ester anc plates t, unless varicm:s zones.
It can be shown by analyses trat location anc shielding of tie oculp-eart reeuces two oose one cose rate.
3.11A-7 ANM en: No. 16
. ~ _ _ _
r.
1 i
(5)
The SENPP model assumes removal by plate-out using a mechanistic model considering elemental, particulate, and organic fractions of halogens I
as vill as the particulate fractions of solid fission products.
l t
The SHKPP model also assumes mechanistic spray renoval and dilution of the 50 percent halogen core inventory and 1 percent solid fission products inventory source terms with the combined volumes of the Reactor Coolant, Accumulators, and the Refueling k'ater Storage Tank.
The resulting sump activity as a function of time is given on FSAR Table 12.2.1-26.
I l
lWIER.T A r
l l
l t
I i-1 l
TABLE 12.2.1-26 SUMP ACTIVITY Energy Group Source Strengths at Time after Release (ys/cc-sec)
MeV/Y 1 Ilour 1 Day 3 Days 7 Days 14 Days 1 Month 1 Year 0.00 - 0.106 4.12(+7) 5.43(+7) 3.89(+7)
~2.51(+7) 1.52(+7) 6.78(+6) 1.65(+6)*
0.106 - 0.440 1.21(+9)
- 8. 58 (+8) 7.06(+8)
- 4. 95 (+ R )
2.74(+8) 7.72(+7) 1.55(+6) 0.440 - 0.865 5.81(+9) 1.23(+9) 3.91(+8) 1.64(+R) 1.26(+R) 8.51(+7) 1.29(+7) 0.865 - 1.332
- 2. 67 (+9) 1.95(+8) 2.12(+7) 3.37(+6) 1.61(+6) 9.42(+5) 1.24(+5) 1.332 - 1.720 1.65(+8) 4.29(+7) 1.48(+7) 2.70(+6) 4.43(+5) 1.86(+5) 1.31(+5) 1.720 - 2.210 7.10(+8) 4.86(+7) 8.38(+5) 1.23(+5) 1.04(+4) 1.90(+2) 2.210 - 2.754 5.42(+7) 5.49(+6) 3.95(+5) 6.98(+4) 3.85(+3) 5.13 1
j
]
2.754 - 3.930 6.83(+6)
- Power of ten i
Peu:3e. 12. 2.- ) - 3 o
ENCLOSURE 6 The following documents activities completed by Applied Physics in response to the IDI, as well as the completion of or commitment to activities that were outstanding at the time of the reinspection exit interview:
1)
SHIELDING CALCULATIONS All the calculations noted in the audit that had deficiencies have been marked " Superseded" and have been replacad by a new set of calculations.
These new calculations were inspr.cted b the IDI team. As a result of the new calculations, the integrated dose te equipment increases in certain areas. Although this increase inay not affect equipment i
qualification since most equipment is 4311ffed to the highest " envelope" dose, Applied Physics has issued a revised document for Equipment Qualification Doses, and the Project EQ team is conducting a formal review of equipment in the affected zones.
Applied Physics is also conducting a review of other calculations that might have been affected by the superseded calculations. This review is ongoing, but so far no affected calculation has been identified.
The revised calculations in some respects use a different methodology than reported in the FSAR. FSAR Change Notices have been issued to revise the applicable sections.
At the time of the reinspection exit interview certain items pertaining to the calculations were considered "open" and these are addressed below:
a.
Deficiency 2.5-1.
New calculation 040 for the Volume Control Tank referenced the FSAR for source terms.
Revision 1 of this calculation has now been issued with the FSAR referenced as a design input deleted, and a reference to the latest Westinghouse source term document added as the design input, b.
Deficiency 2.5-3 New calculation 041 for the TM1 Shielding Design 1
Review was improperly revised by crossing out portions after it had been signed by the verifier. A formal revision 1 to this calculation has been issued, noting previously revised pages and deleting any FSAR references as the design input.
c.
Deficiency 2.5-4 through 2.5-7.
New calculations for Equipment Qualification had not been completed in that Beta daughter contribution had not yet been added and revised EQ Dose Maps had not been issued. The Beta daughter contribution has now been calculated and revised EQ Dose Maps have been issued to project personnel as SHNPP sketches. Affected calculations have been revised to reflect this as well as deletion of FSAR references as design inputs. The assessment of electrical equipment qualification based on the revised Dose Maps is an ongoing effort.
4 l
l 2)
WESTINGHOUSE SOURCE TERMS The inspection of the Volume Control Tank (VCT) calculation indicated that in 1979 Westinghouse issued a revised source term manual. We have redone the VCT calculation using the revised manual; however, there exist i
other calculations performed prior to December 1979 that used the older L
manual for data input. Applied Physics will review all shielding calculations performed prior to 12/79 to see if they used Westinghouse source terms.
If the old Westinghouse source terms were used, the impact L
of the revised manual will be assessed and noted in the calculation.
l Where the source terms have changed, the calculation will be revised.
3)
REVIEW 0F PAST CALCULATIONS Additional Applied Physics calculations have been reviewed to determine if the deficiencies noted in the IDI were typical and to assess the departmental design verification effort. This review was done in two parts; first, a formal design re-verification of selected calculations; secondly, an informal review of a larger number of calculations.
The review was conducted under the direction of the Chief Engineer of Applied Physics by various engineers in the department.
Ten Shearon Harris calculations were selected to cover a cross section of typical Applied Physics work.
These included 5 shielding calculations. 3 Thermal-Hydraulic (dynamic fluid loads, heat transfer) and 2 Applied Mechanics (special stress analyses). A preponderance of shielding calculations was chosen because that was the area noted in the IDI. The calculations selected had been performed at various times from 1976 through 1984, and all were previously verified. The reviewers were directed to do a complete formal check and design re-verification of each calculation to determine if the calculation's conclusions were correct, and to note any deficiencies. A deficiency was defined as any deviation i
from Ebasco procedure E-30 or any technical error.
The reviewers were instructed to pay particular attention to the type of deficiencies noted by the IDI team. However, the effect of the new Westinghouse source term manual noted in item #2 above was to be addressed separately.
4 This re-verification effort found that all 10 calculations had valid resul ts.
No significant deficiencies were found in either the Thermal-Hydraulic or the Applied Mechanics calculations. Some deficiencies were noted in the shielding calculations, mainly in the documentation and use of data input.
However, all the results were determined to be correct. These 10 calculations along with the reviewers i
comments and supporting documentation, such as reviewers independent calculations, were shown to the IDI team during the reinspection.
Recognizing the small sample size of the 10 calculations and the fact they they were restricted to the Shearon Harris project, a second larger review was performed. This review was conducted by the Chief Engineer and the three Applied Physics supervisors.
It covered approximately 100 calculations across all projects. This review was not as rigorous as the re-verification effort described above; but through spot checking, i
examination of supporting documentation and a review of the methodology as presented, it was deemed sufficient to confirm the quality level of the calculation and the design verification. While this review was not without some findings, it did indicate a high quality level of the technical content and that design verification was conscientiously applied.
1
_- - _ -- - ~._,_ _ - -.-
3)
REVIEW OF PAST CALCULATIONS (Cont'd)
As a result of these reviews, the Applied Physics Chief Engineer has concluded the following:
a.
Applied Physics calculations are being pe'rformed in an accurate manner and a workable design verification effort has existed.
b.
The deficiencies noted in the Shearon Harris shielding calculations were not typical of Applied Physics calculations in general.
Corrective action in this specific area was called for which includes a continuing review of the Shearon Harris shielding effort, more direct supervisory participation, and the assignment of a senior radiation protection engineer to the Shearon Harris project.
c.
While the Applied Physics design verification effort was workable, the review has indicated certain weaknesses that must be upgraded and actions must be taken to prevent the deficiencies noted during the IDI from recurring.
Departmental corrective actions are detailed in items 4 and 5 below.
4)
DESIGN VERIFICATION UPGRADE To prevent recurrence of the deficiencies noted during the IDI, a program has been instituted to upgrade the design verification effort.
This program has three aspects:
a.
Formal Training Over the past several months, formal training sessions have been conducted covering those Ebasco procedures that affect AP work.
Most Applied Physics' members have completed these sessions, the exceptions being those members currently on field assignments or otherwise unavailable. These sessions will continue until all members have completed them, and they will be repeated periodically if deemed necessary.
b.
Verification Awareness A series of department directives have been issued and informal meetings and discussions have been held to increase the awareness, by the engineers, of. the requirements inherent in producing verified calculations. The IDI report has been used as an example for the engineers. A heightened awareness among the individuals in the department has been evident as a result of this program; and these measures will continue.
c.
Direct Supervisory Participation Discussions have been held with all Applied Physics supervisors regarding design verification, and they have been directed to become directly involved in the design verification effort by training their group, identifying individuals that need improvement, and directly reviewing completed calculations to assess the quality of the calculation and verification.
The supervisor signs the Shearon Harris calculation as objective evidence of this review and its acceptability for project appifcation. This effort is ongoing.
5)
ADVANCED TECHNOLOGY TECHNICAL REVIEW Due to the highly technical nature of the work done in Applied Physics and other departments under Advanced Technology, Dr R C Iotti, Vice President of Advanced Technology, has announced the formation of a Technical Review Committee. The charter and procedure for this committee are in the formulation stage, however the committee will consist of senior level personnel within Advanced Technology. The committee will have the mission and organizational freedom to assess the technical accuracy or quality of any design document calculation, study or co'1puter code produced in Advanced Technology. The committee will have the freedom to obtain necessary expertise outside the originating department, or outside Ebasco if necessary, to perform its function.
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m ENCLOSURE 7 Ebasco Quality Assurance has recently completed an audit of eleven App 1ted Pitysfes (AP) calculations. This audit was initiated in April of this year to review those AP calculations that were performed to address the concerns identified during the Integrated Design Inspection (IDI). The purpose of the audit was to ensure that the corrective measures initiated in AP, as reflected in the new calculations, satisfactorily address the IDI concerns.
The eleven calculations which were the subject of this audit were of a radiation dose /shfelding nature reflecting the types of calculations reviewed during the IDI.
The audit was conducted in two phases, the first of which was initiated in April and completed in early July of this year.
Ogring this phase, which covered two of the eleven AP calculations, nine concern $ were identiffed.
These concerns were classified into the following five areas:
1.
Unsatisfactory cocipliance with requirements to identify sources of design inputs and assumptions, justification of selection of inputs, and identification of assumptions and inputs that must be confimed as the design progresses (five items).
2 Incomplete docunentation of desf gn review /checkf ng (one item).
3.
Written presentation with characteristics that may not be reproducible (one item),
4, clerical inaccuracy in presentation of calculation (one item),
5.
Documentation not available from official files showing the basis for acceptability of commercial computer program used by AP ($ PAN 4).
These concerns were femally transaitted to the department chfef engineer and group supervisor for corrective action.
The second phase of the audit was conducted in the latter half of July, during which the remaining nine AP calculations were reviewed.
No additional concerns were identified during this period.
In total, the audit covered hundreds of pages of calculations.
The nine concerns identified during the first phase, which have all been corrected by AP, did not affect the technical bases of the calculations.
Taken alone, the results of this first phase indicate that the programs instituted in AP subsequent to the IDI have yielded substantial improverent. The completion of the second phase with ne additional concerns demonstrate that these programs in conjunction with internal audits, modified to include written feedback to the chief engineer and his written response, will result in calculations fully in compliance with industry and corporate quality standards.
ENCLOSURE 8 CIRCUIT AND RELAY CHANGES The following items were closed by the IDI team; however, implementation is subject to inspection by NRC Region II.
Unresolved Item U5.3-1 Deficiency D5.4-1 Deficiency D5.4-2 Deficiency D5.4-3 Deficiency D5.4-4 Deficl+ney D5.5-2 Deficieacy D5.5-3 Deficie ncy D.5.-4 Deficiency D5.6-1 Deficiency D5.7-1 Deficiency D5.8-1 Deficiency D5.9-1 Deficiency D5.10-1 i
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