IR 05000313/1974004
| ML19319E463 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 04/02/1974 |
| From: | Kidd M, Robert Lewis NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19319E438 | List: |
| References | |
| 50-313-74-04, 50-313-74-4, NUDOCS 8004100736 | |
| Download: ML19319E463 (16) | |
Text
.
-
_
~
S
.
.
-
..
.
/ f '..
UNITr O TrATCS t.
, ^ 7[m
.' 4 A
M!C ENEAGY COMM.SZ;ON s./
.s j: ;[. : .
-
- .,, we.yM cy tentju,'T.;;Y C.MEATni!:'
s.
--
r4rc. i o,. n s.,. e r rn *
,
O s,. R
-,..., '
2, i :.
4.
..
r.
- tnei..n...ena i
-
, u,,.,,,,
.,.. 4.-
.
,
,
su j
><
.<.
o.. cia sua
!
RO Inspection Report No. 50-313/74-4
'
Licensee: Arkansas Power and Light Company Sixth and Pine Streets Pine Bluff, Arkansas 71601
.
Facility Name: Arkansas Nuclear One, Unit 1
.
Docket No.:
50-313 License !ki. :
CPPR-57 Category:
B1 Location: Russellville, Arkansas Type of License: B&W, PWR, 2568 Mwt Type of Inapection:
Routine, Announced
,
Dates of Inspection: February 26 - March 1 and March 6-8, 1974
,
Date of Previous Inspection: February 21, 1974 Principal Inspector:
M. S. Kidd, Reactor Inspector Facilities Test and Startup Branch
,
Accompanying Inspectors: None Other Accompanying Parsonnel: None L
4 -/- 7I
'
Principal Inspector:
..(
,.
M. S. Kidd, R'euctor Inspector Date Fa:ilities Test and Startup Branch Reviewed by:
,0. I.
Md
/l/-
'
R. C. Lewis, Acting Branch Chief Date Facilities Test and Startup. Branch
.
l
)
ESIC _
j Q
TIC _
800410 0 73C
.
.
,
.,.-
,.-c
-
-
,. _
.._.-_w-g.
-
.y
..
-, _., -
-
.
__.
.
-..
-- -- -.
-
-
+ - -
.
Y
,
s k
PN
):
RO.Rpt. No. 50-313//4-4 2-
-
,
.
.
.
SUMMARY OF FINDINGS I.
Enforcecent Matters
,
A.
Violations 1.
The following violation of AEC requirements is considered to be of Category II severity:
I a.
Contrary to the requirement of Criterion V of Appendix B
' to 10 CFR 50, tes; procedure 600.03, " Soluble Poison
i Concentration Control Test," was approved and released for testing without containing acceptance criteria against which results of the test would be compared.
The procedure was revised to specify acceptance criteria and was reviewed and approved prior to completion of the inspection.
(Details I, paragraph 2)
I 2.
The following violations of AEC requirements are considered to be of Category III severity:
a.
Contrary to Criterion XVII of Appendix B to 10 CFR 50, test procedure 230.68, " Clean Radwaste System Test," did i
j not provide for documer.tr-ion of certain test data.
A portion of the t
' cn an addendum to the procedure,
,
l was reaccomplished collect and document the required data prior to completion of the int,ection.
(Details I, paragraph 10)
i b.
Contrary to the requirement of 10 CFR 50.55(e), the inability of the four narrow range reactor coolant system -
ptessure transmitters to meet design response times was
=
l ~
not promptly reported to Regulatory Operations (RO),
(Details I, paragra;h 3)
B.
Safety Items None
-
.II.
Licensee Action on Previously Identified Enforcement Matters A. -Violations,
,
The licensee response to the violations enumerated in R0 Report
No. 50-313/74-2, Summary Section I. A, had not been received at
'
the tima of the; inspection; these items will be reviewed on a subsequent-inspection.-
I
,
o
.
f
-
-
w--
w y
=-p m
w-
-
-qm-e't-g F-4'N
--T--*
W wr*r'
E-f'
yw ww-*-r,g--
-
y w
1 r
<
-r-91y-
- --m m
~
.
.
.
$
.
f%
,e
,RO Rpt. No. 50-313/74-4-3-B.
Safety Items There were no previously identified safety matters.
III. New Unresolved Items 74-4/1 Leak in Pumn P36B Recirculation Line Region II was notified by. telecor
'rch 4,1974, of a leak
.
in the high pressure injection pu_. (P36B) recirculacion line flow orifice discovered during hot functional testing. A final report per 10 CFR 50.55(e) is dus April 4,1974.
(Details I, paragraph 4)
74-4 /2 Control Rod Position Indicators Region II was notif.ed by telecon March 4,1974, at erratic behavior experienced on the absolute and relative control rcd drive (CRD) position indicators. A final report per 10 CFR 50.55(e)
is due April 4, 1974.
(Details I, paragraph 5)
'N 74-4/3 Main Steam Line Isolation Valves (ROIF/74-2)
)
'
%
Regulatory Operations Information Request No. 74-2, dated February 20, 1974, requested information concerning potential generic failures of main steam line isolation valves. A response is due March 20, 1974.
(Details I, paragraph 6)
'
IV.
Status of Previously Identified Unresolved Items 72-9/1 Incorporation of Safety Related Equipment in the Final Safety Analysis Raport (FSAR) Q-Lis t
This mat er has been referred to R0 Headquarters and the Directorate of Licensing (DL) for resolution and is considered
'
closed by Region II, (Details I, paragraph 7)
72-12/2 Valve Wall Thickness Verification Program Not inspected.
73-3/1 Completion of Radiological Waste Disposal System Amendment 43 to the Unit 1 FSAR includes a commitment that the solid radwasta system will be operational by October 1, 1974. ThJs item is considered resolved.
(Details I, paragraph 8)
[i.
v
-
.
<
.
,
,
---
,
_-
-
.
.--
--
,
-
-
-. -..
. -, _. - - -..
-
.
.
,
,
O
,
n
,x
.
(,
RO Rpt. No. 50-313/74-4-4-
.
73-5/2 Core Flood System Flow Rate Test The decision has been made by RO and DL that this test will not be necessary for Arkansas Nuclear One, Unit 1.
This itam is closed.
(Details I, paragraph 9)
73-8/1 Procedural Coverage per Regulatory Guide 1 13
'
l Not inspected. See RO Report No. 50-313/74-2, Details I, paragraph 4, for latest information available.
,
73-10/5 Clean Radwaste System Test Procedure
'
Test procedure 230.68 did not provide for the recording of sufficient data to demonstrate all pump capabilities. Ihis data was obtained on a subsequent rerun, therefore the item is considered closed.
(Details I, paragraph 10)
73-10/6 Respiratory Protection Program and Procedures
Not inspected.
73-10/7 Representative Sampling of Gaseous Wastes Not inspected.
73-12/2 Diesel Generator Trips
Not inspected. See R0 Report No. 50-313/74-2, Details I, i
paragraph 6, for latest information available.
73-12/3 Control Rod Trip Test
No change in status.
Comments on TP 330.05 are yet to be resolved.
(Details I, paragraph 11)
73-14/2 Initial Core Loading Procedure
_
Most comments on procedure 1502.04 were resolved by the develop-ment of Revision 1, which was being reviewed by AP&L at the time of the inspection. This item remains open.
(Details I, paragraph 12)
i
~
73-14/3 Leak Testing of the Personnel Hatch
.
.
The Technica1' Specifications regarding testing of the contain-ment personnel and emergency hatches were modified by Amendment 43
['_)%
to agree with Appendix J to 10 CFR 50.
This item is considered l
\\ _,
resolved.
(Details I, paragraph 13)
t I
'
,
u wm
---
,
,
c
--,
--,r
-
"
,
R0 Rpt. No. 50-313/74-4-5-
.;
73-16/1 Radiography Review Not inspected.
73-17/1 Pressurizer Electromatic Relief Valve No change in status. See R0 Report No. 50-313/74-2, Details I, paragraph 10, for lutest information available. This item remains open.
73-17/2 Emergency Operating Procedures Not inspected.
See R0 Report No. 50-313/74-2, Details II, paragraph 2, for status cf licensee effort on these procedures.
73-17/3 Operational Test Program No change in status. This item remains open.
(Details I, paragraph 14)
- 3-18/1 Emergency Planning i
!.
/
Not inspected.
w./
73-18/3 Calibration of Radiation Monitors Not inspected.
73-19/1 inverter Malfunction Not inspected.
,
73-19/2 Reactor Building Ventilation System Ductuork Not inspected.
74-2/1 Vibration / Noise in Reactor Vessel Not inspected.
74-2/2 Control of Temporary Circuit Modifications Not inspected.
'
74-2/3 Controls For Maintenance Activities Not inspected, j
,,
/
\\,
\\
L/
s'
.
e s-
.
.
,
,.
~s
.
I-RO Rpt. No. 50-313/74-4-6-
\\~ sl
V.
Un_t sual Occurrences A.
,Laak in P36B Recirculation Line During hot functional testing (HFT), a small leak was discovered in the recirculation line flow orifice for the Makeup pump (High Pressure Injection) (P36B). The cause of the leak had not been determined at the conclusion of the inspection.
(Details I, paragraph 4)
-
,
.
B.
CRD Position Indicators i
On several occasions during HFI, erratic behavior was experienced on both the absolute and relative indicator systems used to monitor positions of the CRD's.
The problem had not been fully evaluated at the conclusion of the inspection.
(Details I, paragraph 5)
VI.
Other Significant Findings A.
Plant Status p
Hot functional testing for Unit 1 was completed March 6,1974.
(
)
Cooldown of the plant began shortly thereafter with remeval of the N~ s'
reactor vessel head to conduct an inspection of vessel internals and a preoperational (baseline) inservice inspection of the vessel to follow. AP&L's esticate for core loading has been changed from April 13, 1974, to May 1, 1974.
B.
Personnel Changes C. N. Shively joined the ANO staff March 5,1974, after an employ =ent of two years with Sperry Rand Corporation.
He holds a bachelor of science degree in electrical engineering.
Shively is being assigned the duties of Procedure Administrator, vice M. H. Shanbhag.
C.
Controls for Draining of Fuel Transfer Tube
_
Administrative controls and plant designs to preclude compromising containment integrity when the fuel transfer canal and tube are drained were discussed.
It appears that the design coupled with the administrative controls will give reasonable assurance that containment integrity will be maintained.
(Details I, paragraph 16)
-
v
_-..
._ _
.
e
- *..
.
?
\\
- T. H. Cogburn - Nuclear Engineer
- R.
R. Culp - Test Administrator N A. Moore - ManaCer of Quality Assurance J. L. Orlicek - Qwality Control Engineer M. H. Shanbhag - Procedure Administrator B. A. Terwilliger - Operations Supervisor
!
The violations in Section I were discussed.
Information on these items is given in Deta,1s I, paragraphs 2, 3, and 10.
-s The new unresolved items in Section III were discussed briefly.
Y Details are given in Details I, paragraphs 4, 5, and 6.
s_s/
i The status of previously identified unresolved items in Section IV was discussed.
The inspector stated that the items regarding incorporation of additional safety related equipment on the FSAR Q-List, the status of radwaste systems, the core flood system flow rate test, the clean radwaste system test procedure, and leak testing of the personnel and emergency hatches were considered resolved. Details on these items are given in Details I, paragraphs 7, 8, 9, 10, and 13.
He further stated that all other items listed in Section IV would remain open.
_
,
!
a
.
,
- Attended the March 1,1974, meeting only.
'
- Attended the March 8,1974, meeting only.
7s s U
'
-.
.
_
.
-
.
e
'
.
RO Rpt. No. 50-31 /74-4 I-l
-
DETAILS I Prepared By:
r., fl/
8-X-7/
M. S. Kidd, Readtor Inspector Date Facilities Test and Startup B' anch r
Dates of Inspection: February 26 - March 1 and March 6-8, 1974 Reviewed Sy: /Y C.[ A I/d/
S'-
R. C. Lewis, Acting Branch Chief Date Facilities Test and Startup Branch 1.
Persons Contacted
>
In addition to those listed in the Management Interview section of this report, the following persans were contacted:
Arkansas Power and Light Company P. Almond - Reactor Technician L. W. Humphrey - Quality Assurance Engineer G. H. Miller - Assistant Plant Superintendent w
C. N. Shively - Procedure Administrator i
)
Three Shift Supervisors C/
Three Plant Operators Three Assistant Plant Operators One Clerk Chemist and Health Physicist Babcock and Wilcox Company (B&W)
J. A. Bailey - Test Coordinator Bechtel Corporation (Bechtel)
i N. Covington - Startup Engineer 2.
Soluble Poison Concentration Control Test The inspector witnessed portions of the testing conducted per TP 600.03, including boration.of the reactor coolant system (RCS) by feed and bleed, boration of the RCS by batch feed and deboration by feed.
Manipulation of equipment controls, calculations of boric acid and/or condensate additions, and processing of RCS samples to determine
,
boron concentrations were observed by the inspector.
m \\
.
.
g y---
- - -
%ww
,
-
,,,,,
-
--g
_._
_
___. __
_. __
_; -
e D-m
.
,
)
kO.Rpt. No. 50-313/74-4 I-2
.
The purpose of the test was to verify the operability of systems used
,
to change the boron concentration in the RCS and to verify the adequacy of the procedures and calculational methods in operating procedure 1103.04, " Soluble Poison Concentration Control." No major equipment problems were experienced. The calculations in-volving addition of boric acid to increase the boron concentration verified that the formulas in 1103.04 for this function were correct.
During deborations, the final boron concentrations were lower than predicted, requiring a slight change to the formulas in 1103.04.
Testing was accomplished in accordance with the procedures for the functions which the inspector witnessed. In reviewing the test procedure, the inspector noted that it did not contain explicit acceptance criteria against which the results of the test were to be compared. The inspector informed licensee personnel that this omission appeared to be in violation of Criterion V of Appendix B to 10'CFR 50, " Instructions, Procedures, and Drawings." Acceptance criteria were drafted, reviewed by the Plant Safety Committee (PSC)
and Test Working Group (TWG), and addended to the procedure upon the approval of the plant superintendent on February 27, 1974.
g The violation was discussed during the managemer.t interview. The
inspector stated that he had observed that corrective action had been completed prior to the conclusion of the inspection. Licensee personnel stated that the lack of acceptance criteria was an oversight.
3.
RCS Pressure Transmitter A licensee representative informed the inspector at the site on February 27, 1974, of a problem which had been identified on the narrow range RCS pressure transmitters. He stated that B&W had informed AP&L on that date that the transmitters did not meet
specified response times and that the matter should be reported to.
AEC as a possible 10 CFR 50.55(e) item. The inspector was shown a letter from B&W to Bechtel on the matter dated November 27, 1973, which identified the transmitters by instrument number (PT 1021, PT 1023, PT 1038, and Pr1039). The letter indicated that there was a possibility.that a problem existed in that the response times did not meet design values. AP&L quality assurance (QA) personnel received the letter during the first week of December 1973.
.
The inspector noted that the transmitters in question are identified on Bechtel P&ID M-230 as having inputs to the Reactor Protection System. He asked why AP&L had not notified RO:II of the problem in December of 1973 when the letter discussed above was received. Manage-ment stated that AP&L depends upon B&W and Bechtel to make recommendations I
ht V
,
_. _. _ _ _
_.
_ _ _
___
_ __
__._ _
_.
_
.
'
.
.
a,
,
m
'
.
,
RO Rpt. No. 50-313/74-4 I-3
'
.
i j
regarding the reportability of such items and, in this case, no such y
recommendation was made. ' The inspector stated that since AP&L was
,
aware of the problem in December and did not report it at that time,
>
it appeared that prompt riporting of the item required by 10 CFR 50.55(e)
had not been done. He further stated that regardless of the mechanisms
'
.
'
-used to ' determine when a report should be made,. AP&L is responsible j
for prompt reporting.
The matter wa; discussed during the management interview at which i
time the inspector stated that a notice of~ violation for failure to report the item promptly would be issued. Licensee represen-j tatives did not feel that there was a violation. The inspector
-
'
E stated that reporting per 10 CFR 50.55(e) would be discussed
'
further.
- .
,
. 4.
Leak in Pump P36B Recirculation Line
<,
q A licensee representative infonned the inspector by telephone March 4,1974, that a small leak had.been discovered in the high pressure injection (HPI) pump P36B recirculation line flow
orifice and that the item was being reported as a possible i
10 CFR 50.55(e) item. The recirculation lines for each of the
three HPI pumps tie into a header which discharges into the
{
reactor coolant.ptsop seal water return line which passes the water through seal return coolers and then back to the makeup
tank.
Details of the cause of the leak in the flow orifice (FO-1242) had
!
not been determined at the conclusion of the ' inspection. A written
,
report ' on the matter is due ' April 4,1974. The problem was
discussed in the management interview, at which time ~ the inspector
.,
stated that it would be carried as an unresolved item.
'
,
5.
CRD Position Indicators
i In addition to ti,a report discussed in paragraph 4, the licensee representative also reported on March 4, ik74, that on several
,
occasions during HFT, the relative and absolute position indi-
,
cators for the CRD's had not operated properly. Details of the
.
problems were not known, but it was suspected that contacts on
^!
e
the reed-type switches used in the absolute system were stiching
-
i in the closed position and that there was a binding problem 10
'
the stepping motors used to generate signals conitored by the
~
relative position indicating system. The problems were referred
'
to. B&W for resolution. A description of the systems involved is given in Section'7.2.2.3.4 of the Unit.1 FSAR.
!
o
,
i
!'
s
'
,. _. _ ~,_.-,____ -
. _ _
.
&
.
f%
,
'(v} '
R'O Rpt. No. 50-313/74-4 I-4 This item was discussed in the management interview, with the inspector stating that it would be carried as an unresolved item.
6.
Main Steam Line Isolation Valves Regulatory Operations Information Request (ROIF) No. 74-2, "PWR Main Steam Line Isolation Valves," requested information on this type valve in use on Unit 1 and a description of any problems experienced on them. This information is to be used to evaluate possible generic problems. A response is due March 20, 1974. Thia item will be carried as unresolved pending receipt and evaluation of the response.
7.
Incorporation of Safety Related Items in the Unit 1 FSAR 0-List This unresolved item was last discussed in R0 Report No. 50-313/73-8, Details I, paragraph 6.
There were three systems described as being seismic Category I in Safety Guide 29, " Seismic Design Classification," which were not part of the FSAR Q-List. The Q-List defined those systems and components to which the quality assurance program applies.
R0 Headquarters and the Directorate of Licensing gs (L) were made aware of these apparent omissions by memoranda from RO:II. In that RO:II plans no further followup action on the item, k' s_-
'
it is considered closed. This matter was discussed during the management interview. The inspector informed licensee personnel that followup of the matter would be accomplished by L.
,
8.
Completion of Radiological Waste Disposal Systems This unresolved item was initially discussed in R0 Report No. 50-313/
73-3, Details I, paragraph 4.
During that inspection the inspectors were informed that the system for handling spent resins and filters,
.
described by Section 11.1.3.3 of the FSAR, would not be operational at the time of licensing. Amendment 43 of the FSAR contains a commitment that the system will be in operation by October 1,1974.
In that the FSAR states that spent resins can be stored up to one
_ year, the inspector stated that the matter was considered closed.
_
9.
Core Flood System Flow Rate Test This unresolved item was initially discussed in R0 Report No.
50-313/73-S, Details I, paragraph 11.
A position on this type of testing was developed in January 1974 by RO and L which precluded the requirement for_such testing for plants for which the L safety evaluation report had been issued and/or final Advisory Committee
-
on Reactor Safeguards review had been completed. RO:II was informed by letter, dated February 13, 1974, that this testing did rs not apply to Arkansas Nuclear One, Unit 1.
The ' inspector informed (
)
licensee personnel during the management interview that the matter x~
'
was considered closed.
.
.
,
l
- - - -
._
_ _ - _ _
.
.
,
-.
f'~'s RO Rpt. No. 50-313/74-4 I-5 Y)j i
.
.
10. Caseous and Clean Radwaste System Test Procedures This unresolved item was initially discussed in R0 Report No. 50-313/
73-10, Details II, paragraph 6.c.
Comments on test procedures l
232.72, " Caseous Radwaste System Acceptance Test," and 230.68,
" Clean Radwaste System," formed the bases of the unresolved item.
Subsequent review of 232.72 revealed that the procedure was adequately written. This fact was documented in RO Report No. 50-313/73-18, Details I, paragraph 6.
As discussed in R0 Report No. 50/313/74-2, Details I, paragraph 5, changes to 230.68 were made to satisfy the requirements of acceptance criteria 7.4 of the procedure, but there still was a question as to whether the procedure conducted sufficient testing to demonstrate system capability required by acceptance Criterion 7.5.
During the current inspection, discussions with station personnel and review of test data revealed that flow rates and discharge pressures for the clean waste receiver tank transfer pumps P49A and P49B were not recorded for the flow path through demineralizers Tl5B and T15C in seri.
This portion of the test was rerun
.
February 28, 1974, using an addendum to the procedure which provided for recording this data. The inspector informed licensee management 7 ~'s during the management interview on March 8,1974, that a notice of
)
(d violation of Criterion XVII of Appendix B to 10 CFR 50, " Quality Assurance Records," would be issued. The corrective action taken prior to the conclusion of the inspection was reviewed by the inspector.
11. Control Rod Trip Test This unresolved item was initially discussed in RO Report No. 50-313/
73-12, Details I, paragraph 5.
One comment involved performing trip
.
tests an additional 25 times on the slowest and fastest rods found in
'
testing. The inspector informed licensee personnel that this connent was being modified to require only 10 additional trips on the slowest and fastest rods in accordance with Regulatory Guide 1.6.8,
"Preoperational and Initial Startup Test Programs for Water-Cooled
,
Power Reactors," dated November 1973. No action had been taken on other comments, therefore, the item remains open.
12. Initial Core Loading Procedure
.
Comments on procedure 1502.04 were documented in RO Report Nos.
]
50-313/73-14, Details I, paragraph 4; 50-313/74-17, Details I, paragraph 18; and 50-313/74-1, Details I, paragraph 6.
Review of
.
Revision 1 to 1502.04, which was ready for submittal to AP&L review committees, revealed that all previous comments with the
,-~
,
s
s
-
,
,
_
_
.
__ -
..
-.
.
. _ _ ~ _.
. -.
-. -.
---
..
-. - -__-_
>
-
.
-
.
s-
..
'
<s
N :
~ RO Rpt.;No..50-313/74-4 I-6
,
!
L
' exception of response checks of instrumentation had been acted upon in - rewriting the procedure. Regarding instrumentation, the inspector stated that RO:II would expect to see calf brations of the temporary
'
neutron detectors (incore) and source range instruments (excore)
against known sources within a relatively short period of time before t-core loading, such as one to two weeks. Also, as the source range j
instruments are calibrated, the intenmediate range instruments. (IRM)
should also be monitored to determine if a response can be detected by the IRM channels. He also stated that AP&L's proposed method of response checking the temporary and source range detectors when the
first two fuel assemblies (which contain startup sources) are inserted
.
t should be described La the procedure with appropriate instcuccions in t
the event the anticipated responses are not seen. A licensee l
representative stated that provisions covering these comments would be
!
incorporated into the procedure. The licensee representative agreed that the procedure would be reviewed by B&W. This itea remains open.
13. Leak Testing of Personnel and Emergency Hatches
.
l This unresolved item was initially discussed in RO Report No. 50-313/
'
73-14, Details I, paragraph 5.
Amendment 43 to the FSAR modified Technical Specification 4.4.1.2.5(b) to require testing of the outer door seals af ter each opening but no more frequently than daily during
'
,
normal operation or weekly during refueling or cold shutdowns. It also provides for pressure testing the hatches every six months. In that these requirements appear to meet the intent of Appendix J to j
10 (3ut 50, the inspector stated during the management interview that
[
this matter was considered resolved.
l
.
{
14. Operational Test Program h
'This unresolved item w. identified in RO Report No. 50-313/73-17, 1-Details III, paragraph 3.
Discussions with plant personnel revealed i
that for those surveillance tests for which the operations group is responsible, several of the procedures were verified during HFT, but
[
that the procedures to be used by the technical support'and maintenance
~
groups had not been checked out. The inspectors expressed concern
.
that many of these proceduree-would not have been checked out prior -
to their use in completing testir.3 required by Technical Specifications i
af ter' the opersting -license is issued. Also, as discussed in RO i
Report No. 50-313/74-5, Details I, paragraph 4.k, documentation of I
the computer. program to be used in scheduling of tests and documen-tation of master schedules have not been completed. * sis item
[
-remains open.
.
.
$
v
,
i
~
/
[
t
.
-.
-
..... -. - -
-
.
.--
- -
.
-.-,,
., -
.. -
.
s
.
.
n
. R0 Rpt. No. 50-313, /4-5 I-7
.
,--
15. Plant Startup Procedure.
In reviewing general plant procedure 1102.02, " Plant Startup," the inspector noted that there might be a conflict between step 6.4.46 of the procedure (Revision 1) and Technical Specification 3.1.3.5 Step 6.4.46, requires withdrawing one group of safety rods prior to reducing boron concentration (Step 6.4.52) to the point which would give the desired rod configuration during startup. Technical Specification 3.1.3.5, as written, could be interpreted to apply to this operation and it requires withdrawing all four safety groups.
Discussions with AP&L and L personnel resulted in the agreement that the specification did not apply to any operacion except the approach to criticality. The inspector asked if it could be clarified to avoid any future confusion. Licensae personnel indicated that this would be done. The inspector stated that he had no further questions.
,
16. Administrative Controls for Draining of The Fuel Transfer Canal The inspector discucsed controls which AP&L had developed for draining of the fuel transfer canal to assure reactor building containment integrity and to assure that the spent fuel pool water level would not be lowered. The fuel handling systems are described
_
in Section 9.6 of the FSAR. The discussions and review of plant procedures revealed the following information:
Isolation capability for the fuel transfer tube is provided by a.
a manually operated gate valve (SF-45) on the fuel handling area end and a blind flange on the reactor building end, b.
The transfer tube terminates on the spent fuel end in a small pool (tilt pit) containing a fuel assembly upender and other handling equipment. A channel connects this smaller pool to
,
the spent fuel pool. Except for refueling operations, a uater-
'
tight gate is installed between the tilt pit and the spent fuel pool, c.
In the event the fuel transfer canal were drained with the blind flange and water-tight gate removed and the gate valve opened, a decrease in spent fuel pool water level of approximately two feet would actuate an alarm.
d.
With the water-tight gate installed, the minimum spent fuel pool water level is controlled by the fact that the suction line for the spent fuel cooling system pumps terminates well above the
"
,
level of the fuel assemblics.
(See Bechtel P&ID M-235)
s_ -
.
.
,
--.
- - _
.
.
.
.
.
~
.
-
s.
.
20 Rpt. No. 50-313/74-4 I-8 e.
Af ter refueling activities are ccmpleted, and prior to draining of the fuel transfer canal, gate valve SF-45 is closed per
,
procedure 1104.04, " Decay llent Removal System Operating Procedure,"
-
if this system is to be used to drain the canal or by procedure 1102.15, " Filling and Draining Fuel Transfer Canal," if the spent fuel cooling system is to be used for draining (normal
.
' method).
Af ter the canal is drained, the blind flange is installed and its j
gaskets leak tested per procedure 1502.01, " Refueling Operations
'
Sequence of Events."
-
Verification that SF-45 is closed and locked and that the blind flange has been installed is specified by procedure 1102.01,
" Plant Preheatup and Precritical Check," prior to returning the plant to operating status.
The inspector stated that he had no further questions on this matter.
(s /
.
i
,
i
,
'
F G
\\' ^
.
,
e
g
--
- - - - -
-- -
,,,-,
,
,
-,.
-y,.--
-e,,,,-,y
, - -, +-,,,
,-,p-
-r-,
r
.
.
.
~~
.
.
-
s
]
,
RO Inspection Report No. 50-313/74-4 ApR 3 1974
.
ARKANSAS PO'!ER AND LICIIT COMPANY Al;0, Unit 1 DISTRIBUTION:
H. D. Thornburg, R0 M:HQ (5)
'
DR Central Files Regulatory Standards (3)
Directorate of Licensing (13)
cc encl. only:
- PDR
- Local PDR
- NSIC
- DTIE, OR
- State
- To be dispatched at a later date.
s
.
eum e
O