ML19256B813
| ML19256B813 | |
| Person / Time | |
|---|---|
| Site: | Crane, Davis Besse  |
| Issue date: | 04/19/1978 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML19256B808 | List: |
| References | |
| TASK-TF, TASK-TMR 50-346-78-06-01, 50-346-78-6-1, NUDOCS 7908310133 | |
| Download: ML19256B813 (12) | |
Text
/
Appendix A NOTICE OF VIOLATION Toledo Edison Company Docket No. 50-346 Based on the inspection conducted on March 6-8, 1978, it appears that certain of your activities were in noncompliance with NRC require =ents, as noted below. This item is an infraction.
Technical Specification 6.8.1, procedures, states, "*Jritten proce-dures shall be established, implemented and maintained covering the activities referenced below:
- e. Surveillance and test activities of safety related equipment." Technical Specification 6.8.2 states, "Each procedure of 6.8.1, above, and changes thereto, shall be
~
reviewed by the SR3 and approved ey the Station Superintendent prior to i=ple=entation and review periodically as set forth in administra-tive procedures."
a.
Contrary to Technical Specificacion 6.8.1, Administrative Procedure AD 1801.04.3, Section 7.1, was not followed in that when rod drop testing was performed on November 27, 1977, and when the atalysis and review of test results revealed the acceptance criterion was not cet, a deficiency was not docu-
- mented, b.
Contrary to Technical Specification 6.8.2, when rod drop testing was perfor=ed on December 9,1977, changes were made to Procedure 0800.29 without proper review and approval.
% u.
w 1
yh(,u,3in U)iin!hidd-190831453, 1910 179
/,
N O
DETAILS i
1.
Persons Contacted V
- T. Murray, Station Superintendent I
- L.
Stalter, Technical Engineer
- W. Green, Administrative Coordinator J. Lingerfelter, Nuclear and Performance Engineer The ipspector also talked with and interviewed other licensee employees, including members of the technical and operations staffs.
.l
- Denotes those attending the exit interview.
i 2.
Further Review of Reactor-Turbine Trio with Loss of Of fsite Power l
l On Nove=ber 29, 1977, a reactor trip ccptred due to a short in a patch panel used for startup cesting.-
Since the reactor was cooled with the reactor coolant pumps tripped, the licensee de-sired to use data accumulated during the event to support the conclusion that sufficient natural circulation capability exists.
The inspector reviewed charts, data and logs associated with u
the event in order to ascertain whether the data and conditions 9
under which the event occurred allowed an accurate deter =ination of natural circulation capability.
f In the exit interview the inspector stated that based upon his t
review, the data did not adequately support the require =ents of t
the natural circulation test.
Subsequent.. this the licensee
[
presented the data to NRR in a meeting held February 7,1978.
I Sy le:ter from R. S. Boyd to L. E. Roe dated February 16, 1978, NRR infor ed the licensee that the natural circulation would have to be perfor=ed per their coccitment in the FSAR.
This
(
letter did allow a defer =ent of 120 days before the test had to L
be performed.
During the review, the insractor noted that the pressurizer level indication had gone offsca'.e for approximately 5 minutes, and the l
i minimum pressuri:er level was not known. The licensee later I
furnished the i:.spector with a calculation that the level fell approximately 9 inches below the lower level sensing tap.
This j
^t
-n a
i of M.%p,2 9. % g'dh 1/ II Inspection Rpt No. 50-346/77-34.
l wWh
~
cp,Q iu y a yy :uo y
s w 1910 180
O calculation assumed Loop I hot leg te=peratures were the sa=e ds Loop 2 hot leg 'emperatures.
Actual data for Loop I hot leg temperatures w~re not available.
Review of RCS pressure data associated with th reactimeter shosed several pressure variations on the order of 50-100 psi during the event.
The licensee
=aintains the indications are erroneous due to faulty instru-
=entation.
Discussions with a reactor operator and review of control room strip charts *.... to confir= the licensee's position that the readings were faulty.
s No ite=s of nonco=pliance or deviations were identified.
P 3.
Review of Dropped Rod Bank Event On Dece=ber 4, 1977, at 10:33 p.=., safety d gr ups 1 and 3 explanation.27 dropped into the core without Generator output prior to the event was approximately 140 MWe.
Power af:er the rod drop was approxicately 50 MWe.
r At 9:51 p.=., the licensee had successfully co=pleted auto transfer on "A" and "B" safeguards busses si=ultaneously.
This trans fer was acco=panied by a co=puter alar =, "Any Trip Device B/D Tripped."
In addition, the CRD progra==er indicated a fault.
Electronic technicians were i==ediately dispatched to determine the cause
[
of the alar =s, but they could not determine the source.
L The operator recalled seeing safety groups 1 and 3 dropping.
When the safety groups dropped, rod groups 6 and 7 withdrew fro: 26 and 41% withdrawn. Using a rod speed of 30 in/=in, it
(
can be inferred that it took the operator about 40 seconds to take =anual control of the control rods after the safety groups
(
dropped.
F t
It was determined af ter the rod drop that Technical Specifica-tions require that the reactor =ust be placed in Mode 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Generator power was decreased to 10 MWe and the turbine output breakers tripped at apprcximately 11:15 p.=.
Rod groups 5, 6, and 7 were then inserted and the reactor taken into the L
hot shutdown = ode and a shutdown =argin calculation performed.
l V
After the event, the rod control syste= vendor was contacted.
personnel from the co=pany reviewed the event and exa=ined the i
2/ LER Sp 33-77-102, dated 12/27/77.
L r%\\
mys y
3
\\
E'# ~
j@}
y 7
3-
- r.
t P
f m
/
l circuitry, but could not determine the cause for the dropped g
h rods.
No items of noncocpliance or deviations were identified.
4.
Review of Reactivity Coefficient Determination at Power The inspector reviewed data and logs associated with Test No.
TP 800.05.1 which was performed November 29, 1977.
Review of f
the reacti=eter traces revealed reactivity oscillations of a
(
roughly sinusoidal form. The period of the variations was approxi-
=ately 4 seconds and the amplitude was approximately 2.5 pcm.
Reactor power during the test was 40T.*
The licensee attributes the oscillations to fluctuations in steam generator level and to the resulcing~ variations in water te=perature i
and density occurring in the downcomer region of the core. A vendor representative estimated the effect to be less than l'F variation in T cold temperature. This representative states that si=ilar oscillations have been noted at another of the vendor's sites and that the oscillations diminish as power increases.
Because of the oscillations, the reactivity data must be corrected to re=ove the oscillating component.
Figure I shows the behavior of the reactivity trace during the move =ent of control rod groups 6 and 7.
Groups 6 and 7 =ove=ent is ceasured incre=entally by reed t
switch outputs.
During the conduct of the test, the reactor coolant system Tave was 9
l lowered 5'F.
This resulted in BTU limits being received on the steam generators.
Since this was unexpected, the licensee did a t
setpoint calculation for the BTU limit. This analysis showed that f
the value progra= ed into the calculation for reactor coolant flow (67 X 10 lb/hr/ loop) was probably in error. The licensee stated that when full power operation is achieved, the setpoint program I
vill be properly calibrated. In order to avoid the BTU limits problem, the licensee issued a te=porary procedure change to t
lower the Tave setpoint 3*F instead of 5'F.
Further review of t
the test will await data reduction and take place in a future I
inspection. This =atter is considered unresolved.
t t
5.
Incore Detector System Review i
[l A review of TP 800.24, Incore Detector Testing, and ST 5033.03, Incore Instrument Channel Calibration, was performed on Dece=ber 7 and 8, 1977.
Fro this review, several questions have arisen.
j
- Test performed at 75T power with Tave increased 5'F show an al=est complete da= ping of the oscillations.
.o
,. f Nh ep 'e my un-iq' i., g.
ua o.
b 1910 182 t '
g
- 4 W-4 j
q l
i O
Section 7.3 of TP 800.24 under " Acceptance Criteria" states,
/
Normalized pcwer corrected SPND value plots for each string are reasonable and consistent with respect to similar sym:cetrical strings and/or ;eneral flux shapes as deter =ined by NPE per 6.8 (5033.03)." Tnere is so=e uncertainty as to what constitutes
" reasonably.nd consistent", and therefore, whether or not 7.3 (TP S00.24) is defi tive enough to be considered an acceptance criteria.
ST 5033.03 was written and is performed to satisfy Technical Specification 4.3.3.2, calling for a channel calibration which does not include the neutron detectors, to be performed at least once per 18 months. ST 5033.03 is essentially a check of the sof tware used to analyze detector signals and a comparison between detectors located in similar symmetrical strings.
The " Acceptance Criteria" for this comparison is as described in the previous ite=.
ST 5033.03 is not clearly a channel calibration by definition.
Inf ormation concerning the treatment of background detector signals, now deleted from the incore data analysis, was unavailable at the time of this inspection.
Technical information on the aluminum oxide insulated detectors currently in use as part of the incere detector system was unavail-able at the time of this inspection.
Tne s e items are considered unresolved at this time, pending further investigation.
6.
Red Drop Testing at 40: Power The inspector interviewegl personnel and reviewed records associated with rod drop testing at 40: power. The following sequence of events along with cocments was developed.
Pretest meeting notification 10/25/77 Pretest checklist completed 11/26/77 Ch ecklis t items include:
- Procedure review co=plete and approved
- Procedure available
- Pretest deficiency list prepared Q\\\\
ih h \\h s Q w\\
-s p s.[
- e, y \\\\e[
T&qQ
1910 183
.m
)
G
- Pretest deficiencies resolved per Administrative Directive
/
1801.04, Resolution of Test Deficiencies TP 0800.29, Dropped Control Rod Test, run first time.
11/27/77 Procedure Phase I P erequisite 6.2.7 states that the axial power shaping rods vill be positioned to establish a near axial i= balance of near zero.
Actual imbalance was on the order of - 10% at the start of the test.
sPhase II of test not perfor:ed on November 27, 1977)
Procedure Step 7.1.1 states to obtain data specified by the Core Power Distribution Procedure, TP 0800.11, Section 7.0.
No record was found of Enclosure 1, Prerequisite and Procedure Signoff, for this step. This step was verified as completed.
Procedure Step 7.1.8 states to repeat Step 7.1.1.
Again the Enclosure 1 data was not found for this step.
This step was verified as completed.
Procedure Step 7.1.9 states to use Enclosure 1 to compare quadrant power tilt canual calculations with those generated by the computer. for this test was not found. This step was verified as completed.
Procedure Step 7.1.19 states to calculate minimum DNER and the maximum LHR using a copy of Enclosure 2.
A copy of Enclosure 2 was not found.
This step was verified as completed.
The following is an entry in the chronological test log at 1400 hour0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br />s:
" Reviewing the data, we have found that the KW/ft criteria was not met.
The problem appears to be in the initial rod position which resulted in approximately 71% WD position of 6/7 when 5-7 was at 0% WD.
- Ideally, the 6/7 position vould have 78% (100: FP insertion limit),
but due to small rod worth errors, it was not.
The additional 6% on 6/7 vill probably be sufficient to modify the flux distribution to =eet the KW/ft criteria.
The test will be rerun (only the 5-7 rod at 0% WD portion) with an initial rod position of approxi=ately 72% on 6/7 which should put 6/7 at 78% WD when 5-7 is at 0 WD."
n s.
3 d W 6' 1910 184 u
(y \\
A g;y(\\y' s
s
\\\\
s
4 Administrative Procedure AD 1801.04, Resolutien of Test y
Deficiencies, defines deficiencies at "A deviation from an authoriced require =ent document in a component or system identified prior to a test, during a test, or during the review of test results that, were it to remain uncorrected, could adversely affect design and safe operation of the nuclear power plant at any time throughout the expected lifetime of the plant.
The extrapolated value of MLHR to 100% power yielded a value of 20.5 KW/ft. This was in excess of the procedure acceptance criteria of 20.19 KW/ft and the Technical Specifications safety limit basis of 20.4 KW/ft.
Section 7.1 of Administrative Procedure 1801.04 states that if during the analysis of the test data and the review of the test results it is determined that the acceptance criteria were not met, the deficiency shall be documented on Enclosure 1. is entitled
" Deficiency Report." Not filing a Deficiency Report per AD 1801.04 is considered to be an ites of noncocpliance with Technical Spec,ification 6.8.1.c.
Not notifying the NRC per the require =ents of Technical Specification 6.9.1.8.i af ter the extrapolated value of MLHR exceeded 20.4 KW/ft is considered to be an unresolved ites.
12/9/77 - Phase I of TP 0800.29 was rerun.
When the test was perfor=ed, the test leader obtained the shift foreman's permission to run the test.
The test leader checked off selected steps of the procedure to be performed as follows:
Steps 7.1.1, 7.1.2, 7.1.6, 7.1.7, 7.1.8, 7.1.10, 7.1.11, 7.1.16, 7.1.17, 7.1.18, and 6.1.20.
The steps performed were not verified and dated.
Section 6.2, Phase I, Prerequisites, were not verified and dated.
Not even a temporary procedure change accocpanied the perfor=ance of the test to address the procedure =odific-ation. Perfor=ing the test without an adequately reviewed and approved procedure is an apparent item of noncompliance with Technical Specification 6.8.2.
.O sma m
"DI 5b.
' ^
, ' q"c qu'
. ~
S
-1 o
30W
-\\y\\ ;y 1910 185 s
Again, the test results failed the acceptance criteria in
'(
that the maximum linear heat rate extrapolated to 100:
power was 21.1 KW/ft.
Again, there was no notification of the NRC per Technical Specification 6.9.1.8.i.
12/17/77-Test Deficiency Report filed by test leader.
Tes t Deficiency Report states " Extrapolated values of MLHR exceeded fuel celt limit when extrapolated to 100:
FP.
Values acceptable if extrapolated to trip setpoint of 75 plateau (85% FP)."
The recommended action was to request the reactor vendor to review extrapolation techniques to eliminate unnecessary conservatisms.
The responsible section head did not sign the Deficiency Report until February 17, 1978.
The Deficiency Report was not signed off as noted by the plant superintendent as of the date of the inspection.
On December 15, 1977, the test program =anager requested vendor review and co==ents by issuing a Request for
)
Review and Co==ents (No. 186). The request states that the enclosed test docucentation includes no test deficiencies.
A note at the bottom of the request states, however, "This is a preli=inary review only.
The co=plete package for review will be forwarded following resolution of test deficiency." On Dece=ber 16, 1977, the vendor noted in their com=ents that the extrapolation of linear heat rate to 100% full power does not meet the acceptance criteria.
However, an extrapolation to the next overpower trip setpoint (85% of full power) will =eet the acceptance criterion and does not represent a safety concern for testing at the 75: plateau.
The vendor stated that they were presently evaluating conservatism in the extrapolation and would provide the resolution to this test deficiency prior to finishing the 75: testing. Until such resolution was provided, the test deficiency was a restraint to increasing the overpower trip se: point above 851.
'y\\\\
N m'pd v,
,, j
.y'
,N 8-
. gs N
y 1910 186
4 Section 11 of AD 18091.04 states that the test progra=
=anager will =aintain a log of the status of all test deficiencies. As of the date of the inspection, the deficiency associated with MLHR had not been logged by the test progra= =anager.
t on January 20, 1978, the vendor sent =e=o SOM Jo. 335 to TECo personnel which addressed revised hand calculations i
for Fq, F and LHR.
Salient points in the =e=o are as g
follows:
1.
A radial local peak of 1.066 can be used based on our i= proved core =odels.
(SOM 2S3 dated May 23, 1977, had described a radial local peak of 1.10) 2.
The calculation of LHR should use the densification spike factor for the axial level where the peak occurs and not the factor for the 9 foot level.
I Licensee personnel reviewed SOM 335 and requested additienal inf or=ation which war furnished in SOM 336 dated January 31, 1978.
This =e=o explained "The hand calculation of LHR uses conservative factors which add a total of 28 correction to the calculation. Two of these factors (peak to average
(
seg=ent power = 1.04 and the radial local peak - 1.066) l total 10.9% and are picked to conservatively to cover any power distribution.
The online co=puter curve / surface fit routines calculate these factors for the specific power distribution present, and hence, generally calculate nu=bers less than or equal to the hand calculation nu=bers.
l On the first dropped rod test, the online co=puter calculated 3.8% as the co=bined factor of f.hese two i
ite=s; on the second test the co=puter cel:ulated 2.4%.*
l The vendor concluded that the 10.9% factor could be replaced with the afore=entioned values to calculate MLHR.
The resulting calculation yielded a MLHR of 18.84 KW/ft for the test perfor=ed on Nove=ber 27, 1977, and 19.10 KW/ft for the test perfor=ed on Dece=ber 9,1977.
SOM 336 also states future calculations of LHR should continued to use the conservative factors listed SCM 335, and if any further li=its are exceeded, a detailed analysis of that power distribution will also be require'd.
3
[ \\ S QN, v' W gg@>Q' 1910 187
O 7.
Review of Guadrant Power Tilting During Rod Droo Testing Review of online computer calculated tilts at the dropped rod 02 withdrawn condition on Nove=ber 27, 1977, revealed the following:
Quadrant WX XY YZ ZW Incore 13.504 12.474
-9.6816
-16.297 Out-of-Core 1.889 10.233 6.7866
-15.912 Til:9 observed at the 502 withdrawn condition on Nove=ber 27, 1977, were:
Quadrant WX XY YZ ZW Incore 8.6494 7.589
-6.2598
-9.9785 Out-of-Core
.5981 6.2744 4.5303
-11.404 A review of the incore versus out-of-core values for tilt reveal significant differences in quadrants WX and YZ.
The reason for these differences will be examined in a future inspection.
It was also noted that out-of-core channel 8 was reading approxi=ately 10%
lower in power than the actual power level. These =atters are considered unresolved.
8.
Unresolved Ite=s Unresolved ite=s are matters about which more information is required in order to ascertian whether they are acceptable ite=s, ite=s of noncompliance or deviations. Unresolved ite=s are identified in Paragraphs 4, 5, 6, and 7.
9.
Exit In t e rv iews The inspector met with licensee representatives (denoted in Paragraph 1) on December 8,1977, to su==arized the findings of the inspection. The following items were discussed:
a.
Review of the Nove=ber 29, 1977, event with regards to natural circulation capability. The inspector requested that the test be conducted as per the procedures.
(Para-graph 2) p q ra [
73 -,
1910 188
4, b.
Review of the December 4, 1977, safety groups rod drop.
(Paragraph 3)
"s c.
Review of reactivity coefficient at power determination.
(Paragraph 4) d.
Review of incere syste= data. (Faragraph 5)
The following commitments were made by the licensee:
a.
Faulty instrumencation associated with reactor coolant system pressure monitoring will be investigated and repaired if necessary.
b.
An analysis will be provided to deter =ine the minimum pressuri:er level attained during the November 29, 1977, event.
c.
That an analysis of the Nove=ber 29, 1977, event with regards to natural circulations capabilities will be forwarded to NRR for review. This analysis will include a detailed description of the conditions surronding the event.
d.
That further monitoring of the rod control system would be explained in the licensee event report concerning the December 4,1977, event.
On March e, 1978, an additional exit interview was held with the following item discussed:
The performance of rod drop testing at the 40% power plateau.
a.
Lack of notification of NRC regarding MLHR limits.
(Paragraph 6) b.
Lack of filing a Deficiency Report after acceptance criteria was exceeded.
(Paragraph 6) c.
Lack of proper review and approval of procedure used during tes ting which took place on Dece=ber 9,1977.
(Paragraph 6) in reply to the item concerning the notification of the NRC, the licensee stated they felt that the axial i= balance experienced during the test represented an overly conservative condition when compared to conditions that would be experienced at 100%
power. With regard to filing a Deficiency Report after the first test, the licensee agreed that a mistake was made. The licensee also concurred that =anagement controls over the test run on December 9,1977, were not as nor ally exercised and that more emphasis on =anagement controls would be more for exercised for further testing.
.s
.... n -
Attach =ent:
j L,
Reactivity Chart
,s
_. p-.,.p.H v...N w... *..M..... -... 4+..,_k.
-g..-,.m n
~1
- 2. L.. 9 -* +.
~;3..%x &...y. ~.
.. -. -., ~..
--*~~: ?'"~~~~*.~..~*:
e m-e.,, n
. e q p q Q. D f ll r y h s W W.'. 4..
- -.~.,...
3 u. %
.e.
a w - n a....,e...,....,,.~.~...,......,..L.
~.
- s.. ~
4.
.w
....._~-....y...n..._.,.. n v. w >. 9 _.. w. ;,mu...
.. =.m.
. ww u.
- .a.
o o e o e e O O O O O O e O O O O O C C O C O O O O
.. _. _. _ r.c r :u c n. a..:,,,.. : O :..
..n m.
e J.... _.mL. __
_ u _
._._.3..._
__._.J_..
._---u__...
- p- __
4 1 -.
~
j......_._..)._-
. =__.,;
7-
__.m p._.
y.
'_.:.~_p._.t..,.,..._.,a. u..,__....
,...4
,.. _ w...
.. ~..
_t
. _. _ _. _. _.. _.._._J.._
y _
4 l
q
___1
. _ _ s.. -
______..q.-
. m. __
i.
.O.
__.__._.N_
__.. _.i ___ ___
._. ~$, ___._._..O.- y -
t
- s. __..
,i i
i t
r--
t 8
[
____ y 3
g
.t--------
.._. _ q -
.O._
-H.
^
g i
_._1.___._.. O i
,s.
_d_--
4 3-
__7..
___ _ _- O.
- __a, 3,
_j.__;o
~_
_l__
i
- p. -
s
- __... __. r,
_q---
..f
. _._ _g g
y D.- /\\---/b r \\
-/S
,.A-A
.'i -r'v---th -t r-T_- -._
4 _.._ _
a.
_M_.__.H. -_ O ;. PW-(-f-- i !
_\\-/
'g... -(/ y..r --W-i
.+--.
9 4---
i
<a -1 s
1--- -- - Oy
/>
' _m-.f-M_; _-
m
_v f.
t
,,, 7, m
..p..g i
i y_
i r-v, y
}_
.._o.
i
. g.__.__ __ C._ _.J i
_j.. _p o
_ O __.. _ _____ -
N I
I
.i a
p.
['_..._,
m
_.i-1 7
q
__i Q-Q J
.._. O
_Q I
I
.O I
1 ON I
l t
8 l
6 I
.u s.
__q
.1 _
... q'. -
.a.
,i_-_._
_ 4
-.. O 2
_~
O
,j
__ _ i --
e 4._.
.__i..
a 6
A..
t I
.me_ _.
- _3
_4
=_
.+
_._s.,____.
- _g,n.. -
.... ^
8 w_..__-
}-
f
.._3____.,,-
- 7.,
, _g__
____.O_..
7
_ 4.-
pg __--
.s._
_ L,. _ _1
.... _.. _, O __
t
__1 J
4 _. _ __..._...-
- _.. -