IR 05000155/1975011
| ML20002D767 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 09/05/1975 |
| From: | Hunter D, Jordan E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20002D764 | List: |
| References | |
| 50-155-75-11, NUDOCS 8101220432 | |
| Download: ML20002D767 (23) | |
Text
,
.-
-
UNITED STATES NUCLEAR REGULATORY C01D11SSION OFFICE OF INSPECTION AND EhTORCEMENT.
h
- REGION III Report of Operations Inspection
~
i
,
IE Inspection Report No. 050-155/75-11
.
Licensee:
Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49210 Big Rock Point Nuclear Plant-License No. DPR-6'
Charlevoix, Michigan Category: C
.
Type of Licensee:
.
Type of Inspection:
Routine, Announced
'
Dates of Inspection:
August 11 - 15, 1975
[
J t-
~7.>T Principal Inspector:
nD.
Hunter
.
'(Date)
.
_
Accompanying Inspector:
E. L. Jordan
.
8/12 & 13/75
,
Other Accompanying Personnel: None Reviewed By:
da
7;~
Seni Inspector
' (Date)
Projects Unit 2 Reactor Operations Eranch
.
I i
I i
r 310/44fR
.
a w
---.*,---g-
=g yw. - - -3-
--w, y
y w9e c
9-e yyy--
g v-v ym y r-
-=
+g a+-m
,y-rm
..
.
SIMfARY 0F FINDINGS.
-
Inspection Summary
' Inspection of August 11-15,- 1975, (75-11): ' Abnormal Occurrence Reports, Review and Audits, Selected-Outstanding Items,' Selected Inspector Identified Items, Review of Safety Limits, Limiting Safety Systems
~
Settings, and Limiting Conditions ^for Operation, and a-Facility Tour.
Enforcement Items
.
~
~
The following item of nonco=pliance was identified during'the inspection.
A.
Violations
.
None.
B.
Infractions-
.
None.
C.
Deficiencies None.
Licensee Action on Previously Identified Enforcement Matters None.
Other Significant Findings A.
Systems and Components Unresolved Item:
The operation of the containment personnel hatch and equipment hatch doors and equalizing valves does not appear to assure the establishment of two individual boundaries (both doors -
and respective equalizing valves closed) after normal ingress or egress evolutions.
(Report Details, Paragraph 7)
,
f
.
-2-it
.
-e+e t'
.
,
ns
-
,
-
.
B.
Facility Items-(Plans and Procedures)
' Unresolved Item: The review and the scope of'the' review of plant operations by the plant freview committee was not apparent.-
(Report Details, Paragraph'2)
'
C.
Managerial Items
-
None.
D.
Noncompliance Identified and Corrected by Licensee
1.
Contrary to Criterion XVI.of' Appendix B.to 10 CFR 50, the-failure to identify and correct a procedural deficiency
>
related.to quality resulted in the failure to meet the-requirements of Technical' Specification..-This item is an infraction.
(Report-Details, Paragraph ~1.1)
2.
Contrary to Technical Specifications 6.4.3.D, the discharge canal grab sample was not obtained and analyced as required.
This item is"an infraction. -(Report Details, Paragraph 1.k)-
E.
Deviations None.
.
F.
Status of Previously Reported Unresolved Items j
The Type C testing of the containment ventilation supply valve (CV-4097) with the' test pressure applied in the opposite direc-tion to that under the design basis accident is acceptable.
(Report Details, Paragraph 6.b)
Management Interviev
.
The management interview was conducted on August 15, 1975, with the following persons present.
C. J. Hartman, Plant Superintendent G. C. Tyson, Maintenance Engineer
R. E. Schrader, I6C Supervisor R. V. Voll, Reactor Engineer
.
J. J. Zabritski, Quality Assurance Engineer
,
-3-i
.
-
JA.
Review of Abnormal Occurr7nce R*porto
.
Tha insp;cter ctctcd that the raview of cbn:rmal eccurr:nces
~
revealed ~that six occurrences had been resolved but the corrective actions for a number of abnormal ~ occurrences reviewed were not completed.
B.
Review and Audits
_
The inspector stated that the review and audits included the respoasibility of the Plant Review Committee review function,-
.
the S.fety Audit Review Board: review, and the Quality Assurance-review and audit function. The inspector noted that tbc. review included selected items.
(Report Details, Paragraph 2)
~
1.
Review of Proposed Technical Specification changes by the-
.
.
2.
Review of Enforcement Items by the PRC.
3.
Review of plant cperations by the PRC. The inspector stated-
that the review of plant operations by the PRC was not apparent
and would be carried as'an unresolved item.
4.
Review of PRC minutes' by the SARB.
5.
Quality Assurance Review and Audit of plant operations. The inspector noted that the QA group had audited the plant operations implementation of the Quality Assurance requirements.
C.
The inspector asked the licensee to consider procedural revisions-or other measures to improve visibility to reactivity increases and reduce' incremental notch worths during future shutdowns margin
,
tests.
D.
The inspector noted that a review of the. performance of the last.
two RT 'hr Procedures by the reactor engineer revealed no apparent h
discrepancies.--(Report Details, Paragraph 4)
E.
The inspector stated the review of the reactor'startup of June 7, 1975, indicated a discrepancy between the computer calculated
.
estimated critical rod position from Jackson and the hand calcul-ated estimated critical position made by the Reactor Engineer at the site; and the inspector noted that the discrepancy was not resolved with management prior to continuing the' reactor startup.
The' inspector stated that the procedures at the plant required-4_
(
t
.
.
-
.
,
management-rzview of nuclcar ccfaty-related discrzp nci:2. The inspector stated that this was an item of noncompliance on the
. previous inspection (75-10). The licensee acknowledged the above statements.
(Report Details, Paragraph 5)
F.
The' inspector stated the NRC position on one'of the~three unresolved-items in IR 75-05 was as follows:
1.
Containment testing as required by Technical Sp2cification 3.7
' to meet the requirements of Appendix J to 10 CFR 50, in the-specific instances of the containment side, gasketed flange on the 24 inch ventilation supply valve (CV-4097) and the emergency escape lock will be continued as unresolved items pending.
the Big Rock plant response to the request from licensing-regarding Appendix J.-
The testing of the containment supply ventilation valve-(CV-4097) in the opposite. direction to that of the flow under the design basis accident is acceptable
,
due to the design of the system.
(Report Details, Paragraph 6)-
'
2.
The inspector stated that the design: adequacy of the vacuum relief system is under study and evaluation by licensing.
C.
The inspector stated that a review of the operation of the contain-
.
ment personnel access hatch, the equipment access hatch, and the.
emergency escape lock revealed an apparent discrepancy relating to the personnel and equipment hatches equalizing valves. The inspector requested procedural control to be placed on the operation of the doors equalizing valves and a subsequent engineering evaluation to be performed to consider a permanent design change.to insure proper operation of the equalizing valves mechanical interlock.
system. The licensee acknowledged the above statement and request.
~
The inspector stated that this would be carried as an unresolved item.
(Report Details, Paragraph 7)
H.
. Review of Safety Limits, Limiting Safety System Settings, and Limiting Conditions for Operation.
The inspector noted that the review of the plant status during selected maintenance, operating
-
and shutdown activities indicated no apparent discrepancies.
(Report Details, Paragraph 8)
I.
The inspector noted that the performance of the monthly surveillance-test T-30-14 (Core Spray Heat Exchanged Leak Rate Test) had not-been performed during the months of February, March, and April, 1975.
The test was performed on June 2, ~ 1975, prior to unit startup on June 5, 1975. The inspector stated that no documentation was-apparent
.
-5-
,
.
..
-
.
,
' which indicated th2 cenigement rsview'cnd cuthoriz: tion.to eliminste the testing. The licensec stated that the plant was in the Cold Shutdown condition while performing the corrective actions associated with A0 50-155/75-01.
(Report Details,-Paragraph 9)-
l(
J.
The inspector stated that a tour of' the facility revealed a number of'important items:
,
1.
The general plant' areas ' vere clean.
2.
The radiation zones were localized and appeared neat and clean.
3.
There was no apparent majorequipment Icakage.
4.
The combustible materials had been -removed from the storage room 441 5.
The inspector noted that-the condensate storage tank vented
-
directly.to the environment. The licensee's' representative
.
indicated the question.would be reviewed.
6.
The inspector noted that'the conduit to the emergency escape lock outer door interferred with the outside inner
door operating mechanism.
7.
The inspector stated that the' station battery which supplies
engineered safeguards systems had no vertical and lateral supports. The Licensce's representative indicated the j
,
question would be reviewed.
The licensec acknowledged the above comments and statements.
(Report Details, Paragraph 10)
The inspector noted that he observed the weekly testing of the K.
emergency dicsci generator. '(Report Details, Paragraph 11).
'
The inspector. requested the licensee to provide an update of the L.
status of the corrective actions presently implemented and planned
-
concerning IE Bulletins No. 74-04 and No.74-04A.
This request r
!
included:
'
1.
Temporary measures taken during the interim while fire bar-riers are not installed.
(Response by August 25, 1975). The
,
licensee stated that the interim ucasures are as indicated in the CP to IE:III response to IE Bulletin No. 75-04 and No.75-04A dated April :27,1975, June 26,1975 and JulyR29, 1975. The licensee stated that contact had been made with the local-6-f
.
W l
-
Y t
l i
.
-.
..
..
.
-
. -
-.-..
. _..
-
.
-
.
.
fire protection agency and the normally scheduled fire
_
protection training-is_ scheduled for September,-1975 (Consumers' retrains plant personnel every 2-years).
' '
2.
Status of completion of the fire barrier fix..(ResponscLby
~
August 8, 1975)..The licensce: stated that-the' status was the same as indicated by Consumers on; July.29, 1975, response-update to IE Eulletin No. 75-04 and No.75-04A. The licensee stated that it appeared the fire barrierLfix would be per-
,
formed (as determined) during the January 3976, outage at simultaneously with the RDS modification; and the fix most likely would include cable and penetration. barrier placements and additional automatic fire and smoke detection equipment.
~
3.
Status of response to Bulletin-No.75-04A (Response by September 2, 1975).
The licensee s'tated:that the answer
to items 2 and 3 would apparently be covered by the same
-
response.
4.
Measures or special measures to control access to the compressor room and the cable spreading room. The licensee stated that the normal access control exists for -the compressor. room and the cable spreading which is the plant procedure for all vital areas.
.
.
S f-
.
7-
-
<
.
.
+
,~g
.. -
,
p-.-
-.e
=
%
p e -p
-
r 1s
-
g-
+
g er
-
'
.
.
'
.
REPORT DETAILS
.
Persons Contacted
-
,
,-
.C. J. Hartman, Plant Superintendent
'D. E..DcMoor, Technical Engineer
,
G. C. Tyson, Maintenance Engineer.
._
)
S. A. Axtell, Chemistry and Radiation. Protection Supervisor R. E. Schrader, Instrument and Control Supervisor V.'A. Avery, Shift Supervisor J. L..Keumin, Associate Engineer-
,
R. A. McVay, Maintenance Supervisor R. W. Voll, Reactor Engineer.
.
.
J. J. Zabritski, Quality Assurance Engineer
1.
Review of Abnormal Occurrence Rep' orts The following abnormal occurrence reports were reviewed to ascertain that the reviews, evaluations, information and corrective actions were as reported to plant management and the NRC.
-
a.
A0 50-155/02-75, Uncontrolled release of 'iquid tgjthe discharge
canal on January 22, 1975. The licensee.
that approximately 50 gallons of lou-activity"paported-feedwater was dumped to a turbine building floor drain by a plant repairman during a maintenance evolution. The inspector verified the document-ation of the investigation of.all turbine building floor drains
-
to determine the release paths and the installation of local name-plates (warning signs) at the individual drains.1 A number of these drains were viewed during the facility tour.
The inspector also noted that one floor drain outside the-containment spray room (427) vas not marked.
(Report Letails, Paragraph 10.j)
The inspcctor-also verified that the maintenance crew had been
~
made avar
~ this particular occurrence by management'and appropria training was performed.
b.
A0 50-155/0v 7ailure of off-gas sample, drain valve to close on Janus 16, 1975. The. licensee reported that the operational test, UES-3, conducted on January 16, 1975, had resulted in certain anomalics igjthe data, which indicated a possible off-gas system leak.--
Subsequent testing on 1/
Ltr, CP to DL, dtd-1/31/75.
2,/
IR 50-155/75-09.
3/
IR 50-155/75-08-8-f
-
.
-
-
,
Fabru:ry 12, 1975, rsv:aled thtt drain valvh SV-RL-27citaked to the turbine building sump, at which time the. incident was.
reported. The inspector verified the replacement ofsthe solenoid. valve (SV-RL-27).with a new' solenoid valve which
-
was mounted in the vertical position (as recommended by the
~
~
manufacturer) and the modification of 'the circuitry-to provide for the valve to - fail closed upon loss of power.
The inspector reviewed-the modification packageJ(Facility Change FC-290) which was. completed and reviewed on June 6, 1975. The final print changes were completed on June 12, 1975.
c.
A0 050-155/08-75, containment ventilation supplyp valve-exc.essive leakage on March 18,-1975.
The licensee, reported 4/
that during the semi-annual testing of the containment supply-ventilation valves (CV-4096 and CV-4097), the IcaKage,(1428 sibs /-
day) was'in excess.of the Technical Specifications:and the Appendix J of 10 CFR 50 limits (.25 percent / day (445 lbs/ day)
~
and 60% of.25 percent / day (267 lbs/ day) of the containmente atmosphere for all co=ponents respectively).
,
~The inspector reviewed the valve seat adjustment by the manufacturer at the facility and the subsequent test pro-cedure performance af ter the adjustment was completed (T 180-01),
'
which returned the total individual component-leakage to 180.32, pounds / day (60% c.f 25 percent / day) less than. The
.
.
inspector verified that the monthly leak rate tests.are being performed during the interin while the review of the valve 3,,
seat material is being completed to determine the shelf life, ed service life, and the effect of the environment on the scat-
'
material.
t d.
'A0 050-15;/14-75, Containment ventilation ^ supply valve
,7
~
d excessive leakage on May 26, 1975. The licensee reportcd that during special monthly test of the containment ventila-tion supply valve (CV-4097), the actuation mechanism bolted fitting had shifted and the valve dise did not rotate to its fully closed position. The inspector reviewed the occurrence-
~
with the licensce's representative and noted that the repair
-
and' punch mark appeared to be adequate. The operating mechanism fitting has not shifted during subsequent monthly
.
-
surveillance testing of the valve.
c.
A0 050-155/09-75, Emergencydieselgeneragprteststartfailure on April 10, 1975. The licensee reported-that during the performance of the routine weekly test (T30-13), following 4/
Ltr,.CP to DL, dtd 3/31/75.
5/
Ltr, CP to DL, dtd~6/5/75.
6/
Ltr, CP to DL, dtd 4/21/75.
-9_
i(-
-
t-e
,
p
_ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _.
_ _ _ _ _ - _.
_
_
._.
_ _ _ _
_____ __
. _ _ _ _ _ _ _.
.
.
h:nd priming,'tha diccol atnrtad nsrm:11y. -Tho in:pector verified that thz lic:n;;c waa continuing cn investigativa program, and the inspector noted that the_ diesel had been.
difficult to' start and operate on June 26, 1975, during ihe
.
performance'of.the routine weekly test..The'licensec's-representative indicated that the sluggish starting evolution
appeared to be similar to the event on April 10, 1975. The inspector verified that the diesel had notf ailed to start f
,
after hand priming of the fuel. system.
f.
A0 050-155/11-75, Pressure switch drift on condenser-low The licensec reported 77 vacuum scram interlock on May-20, 1975.
s that during a routine calibration of the vacuum-interlock ~
switch:(PS/RE ISB), the.setpoint was-found to be at 362 psig.
The inspector verified the corrective action and n'otsd IEnt
a change to Technical Specifications had been' requested in-order to widen the operating band on the switch'(including the; other three identical switches, PS/RE'15A,.15C, and ISD).. The
. inspector also noted that the switches.were wide range (3200-psi) and a study by the licensee indicates that the switches
. would function adequately.within the newly requested'opeidting band. (< 500 psig vs ( 350 psig).
g.
A0 050-155/12-75, Coresprayvalvesgperatorlocknutlooseon day 21, 1975. The licensee reported-that during testing of MO-7051 and M0-7061 the locking nut on the drive mechanisms had loosened several turns. The valves-did not fail to operate, but the potential for failure existed.
The inspector verified the-documentation of the staking certification of the five SMA-00 valves in two positions (nut
.
to threads) 180 apart, and the inspection after testing.to.
~
insure no subsequent movement (MGP-08, Revision 0,_Limitorque Valve Nut Inspection, performed-on June 4,1975). -The inspector noted that the manufacturer made two recommendations for a fix of the problem (staking or set screws). Due to the difficulty.
of drilling and tapping the nut and shaft in order to install the set screws, the staking operation was selected and performed.
The licensee and the manufacturer' indicated that with proper staking, the locknut should not loosen.
.
7/
Ltr, CP to DL, dtd 5/30/75.
8/
Ltr, CP to DL, dtd 6/2/75.
- 10 -
i-k'
T
'
.
.
- ~
w y
.
.
h.
.A0 050-155/13-75, Emergencycondenseroutictivalvc-opgyator
-
locknut loose on May 23, 1975. The licensee reported-that:
during the inspection'of the Emergency, Condenser outict'
valve operator (MO-7053) as a result of'A0 12-75, the lock-nut _was tightened and staked.
The inspector verified the documentation of the staking'ccrt-ification of the five SMA-00 valves in two positions-(nut to-threads) 180 apart, and the inspection after testing to insure no subsequent movement (MGP-8, Revision 'O, Limitorque -
~
Valve Nut Inspection, performed on June 4,1975)..The inspector; noted that the manufacturer made two recommendations for a fix of the problem (staking or set screws). Due to th'e difficulty of.the drilling and tapping the nut and shaft in order to install the set screws, the staking operation was selected and performed. -The licensee and the manufacturer indicated that with proper staking, the locknut should not loosen?
i.
A0 050-155/15-75, Containment isolation valve' leakage during licenseereportediOyalveCV-4027).onMay-30,1975.
test (reactor dral The
-- that during routine testing-of contain-ment valve _CV-4027 (fuel pool and reactor l drain valve), the valve leaked excessive;y. The inspector verified that the valve had been adjusted and retested properly. The procedure (MCP-
-
4, Revision 0) for adjusting the valve travel had been used-in q'
April of 1974, with hand written changes noted.
It was noted.
by the licensees' representative and the inspector that the
^
"'
procedure had been used incorrectly;at the time, resulting'in j
the valve (CV-4027) not being properly adjusted, and it had been
'
- utilized again for adjustments on May 30, 1975.
It was noted that the procedure was not adequate during a Quality _
-Assurance Audit of Safety-Related maintenance procedures on June 13, 1975. The failure to promptly identify and correct the procedural inadequacy (Step 5.8 of MGP-4) in April of 1974, resulting in the failure to meet the Technical-Specification is considered an item of noncompliance pursuant to Criterien XVI ^
of Appendix B to 10 CFR 50.
The discovery of the procedut !
inadequacy was through the program impicmented by the licensee and the appropriate corrective actions have been taken (the procedure was invalidated to; prevent recurrence).
j.
A0 050-155/16-75, Procedural deficiencies associated with the.
reactor and fuel pit valve (CV-4027) inspection and repair on thy 30,1975 (classified as an A0 on July 8,1975).
.
9/
Ltr, CP to DL, dtd 6/2/75.
lj0/ Ltr, CP to DL, dtd 6/10/75.
- 11 -
e
.
-
.
that c Qutlity Accurance Audit cf'cifcty-
~
The licsnets repsrtad related maintenance activitics on June:13, 1975 and subsequent
. investigation and review by the Plant 1 Review Committee on July 8, 1975,-identified the procedural inadequacies associated with-MGP-4f Revision 0,.as generic in nature. The procedure did not'
provide adequate' detail.to perform the-required valveJadjustments.
The inspector noted that the maintenance.procedu,c (MFP-4; r
Revision 0) had been invalidated and previous uses of the' procedure
_
had been identified and evaluated to-insure all associated valves _
had been subsequently leak tested. The licensee is in the' process of reviewing the subject of genera 11 maintenance procedures:
-
and. specific maintenance-procedures to determine _the appropriate long-term corrective action.
In the interim all general, main-tenance ptocedures are being reviewed to_ determine applicability and adequacy. The inspector-verified that:similar safety-related maintenance would not be performed by the licensee unless the general procedure had been reviewed for adequacy or a new approved.
~
procedure had been-provided.
k. ~ A0 050-155/17-75, Discharge canal water not. analyzed on' July 18, 1975 (c1 1fied.as an A0 on July 23, 1975).- The licensee reported 997 :that the daily discharge canal water sample was not
-
taken on July 18, 1975, due to a technician failing to perform the associated ite= on'a daily checklist.
The inspector verified that the daily check had been initiated as stated by the licensee.._This check is. performed by'the Lab technician informing the shift supervisor that the specific:
chemistry and radiological sample are completed. The shift supervisor records this notification on the operations group
,
surveillance status board on a daily basis. The inspector also noted that the actual results of the routine chemistry surveillance (check sheets) are posted in the chemistry lab-oratory window for viewing by cognizant personnci as needed.
The failure to obtain the daily discharge canal grab sample is an item of noncompliance pursuant to the Technical Specification 6.4.3.D.
The discovery of the missed surveillance was through a review program impicmented by the licensee; and the appro-a priatecorrections have been taken to prevent. recurrence, including improvements in the review program by supervision to assure that daily surveillance is reviewed each day.
1.
A0 050-155/01-74, Failureofjg7 Ch9ngg)3neutronfluxlevel instrument on March 1, 1974.
The inspector reviewed the status of the modification to the neutron flux Icvel instrument 11/ Ltr, CP to DL, dtd 7/17/75.
12/ Ltr, CP to DL, dtd 7/30/75.
1R 050-155/74-03.
14/ 1R 050-155/75-08 3 / Ltr, CP to DL dtd 3/12/74.
- 12 -
"
,
.
.,
.
m
- e s-
- r
--
9-e-
.
.
,
Emplifinre to pr:clude similar feiluraa'. The. inspretor noted
. ;
that.the engineering was being dons by; tha vendor end the major.
' '
prob 1 cms appeared.to be the age of the equipment and the hard-vare requirements; but the modification of the amplifiers should be available soon. The inspector verified the recent contact
'i-by the licensce's representative with the Cencral~ Electric engineering group. 'Due to the~ apparent generic nature of this-
.
. occurrence, the inspector pointed out.the importance ofLobtain-ing the recommended modification and determine if further action.
is required by.the vendor,'the licensee, or the Nuclear Regulatorf Commission.
a AO 050-155/23-74, Emergency Lock Test Fixture left installed following testing on September 17 - 1974' (penetrations test -
a~d.theemergencygyegpplock performed on April-29, 1974, n
inner door had been open since June.21, 1975)
The inspector reviewed the corrective' actions associated w th the occurrence to assure that.all the test procedures;had been reviewed as indicated by.the licensee:
.
(1) - PRC meeting 36-74 (November 15, 1974) documented revieu-of procedures by the Operations. group,. radiation protection group, and engineering group completed.
(2) PRC meeting 42-74 (December 2, 1974) documented revicu of'
procedures by the maintenance group completed.
(3) PRC meeting 21-75 - (June 2, 5' and 6, ~1975) ~1ocumented review of procedures by the instrument departscac. - The inspector audited the I&C reviev list and noted that approxinately
,
26 of the 33 procedures required changes.
,
A 2.
Revicu and Audits
'
The following items were reviewed by the inspector to verify that the requirements of the Technical Specification were being met.
The inspector verified that the Plant Review Committee was
a.
held at monthly intervals or'more frequently for the period of July, 1974, through July 28, 1975. No apparent discrepancies were noted.
b.
The inspector verified that the Plant Review Co=mittee meetings.
were held during the period of July,1974, through July 28, 1975 met the quorum requirements. No apparent discrepancies were noted.
.
16/ Ltr, CP to DL, dtd 9/27/74.
17/ IR 050-155/74-10.
- 13 -
.
~
l:
.
.
cetor verified that thi Plcnt Rr. view Crmmitter rd-
.Theins[heAbnormalOccurrencereports'for1974and1975 c.
viewed.
through A0 17-75 which was~ reviewed at'the'PRC meeting'24-75 held on July 28, 1975. No apparent discrepancies were'noted,
,
d.-
.The inspector verifiSd that the' Plant Review Committec; reviewed'
selected items of noncompliance associated'with NRC inspections.
'
.(1)
Inspection Report 74-10,18/ PRC meetings from September, 1974, through June.6, 1975,-(PRC 21-75) when the pro-
- cedures reviews were completed.
.
'(2)f Inspection Repor't 75-04,E PRC meeting on June 6, 1975, (PRC 21-75).
(3) Inspection Report 75-05,E PRC meeting on March 3, 1975, (PRC'8-75).
No apparent discrepancies were noted.
.
The inspector verified that the Plant Review Committee reviewed c.
selected proposed Technical Specification change's:
(1) Technical Specification change 40, Setpoint changes for two' criticality monitors (PRC 6-74).
(2) Technical Specification change 43, Removal of CRD rollers from 16' peripheral' rods (PRC 25-74).
(3) Technical Specification change 44, MAPLIICR (PRC 26074).
,
(4) - Technical Specification-change 45, High Energy pipe surveillance (PRC 42-74).
,
No apparent discrepancies were noted.
f.
The-inspector reviewed the Safety Audit and Review Board ninutes for the period of April 26, 1974, through March 14, 1975, to
-
verify the regulatory meetings were held at quarterly intervals
or more frequently. No apparent discrepancies were noted.
g.
The inspector verified that the Safety Audit and Review Board held during the period of April 26, 1974 through March 14, 1975, met the quorum requirements. - No appaient' discrepancies were noted.
18/ Ltr, IE:1II to CP, dtd 3/11/75.
3/ Ltr, IE:III to CP, dtd 5/2/75.
M/ Ltr, IE;III to CP, dtd 5/20/75.
.
- 14 -
' t
,
!
'
.
y
-
y
,
e v-
-
-
r--
,
r
.
.
h.,
-Ths' inspector ncted.the following iudita w:ra reviiwsd by; the Safety Audit and Ravitw Botrd:
(1) L March 1975, QA.- BU - 009 ' Reactor Engineering
_ (November 2, 1973)
I
~
(2) March 1975, QA - BU- 007 Health Physics (May 32, 1974).
.
.
1.
The inspector review of the Plant 1 Review' Committee requirement.
to review operations to detect potentia 1' safety hazards in
.accordance with Administrative Procedure 1.3.A.3.1.(f) and-the proposed Technical Specification section 6.5.1.6.f was not apparent.
-
.
(1) The-inspector verified that selected operating records had-been reviewed by key staff personnel during the period of July 1, 1974, through~ July 22, 1975; however the-QA-05 form did not-indicate that the PRC review was required and during-the review of the PRC minutes, it was not apparent that the'
,
review of operations was directly documented.
-(2)
It was not apparent to the inspector subsequent to review
'
of the Technical Specification section 6.5.1.6.f., Administra--
tive Procedures 1.2.C.2.1,J1.3.A.3.1(f), 1.14.A.2, and 1.17 what items, activities, and checks are included in the scope of the " review'of plant operations of potential safety hazards".
(3) The apparent lack of procedural Euidelines established to-
.
comply with this_ Technical Specification requirement will
'
be carried as an unresolved.iten, j.
The inspector also noted the following.special items whichLwere reviewed by the Safety Review and Audit Board.
(1) Plant Review Committee minutes:
SARB meeting on July 15, 1974 SARB meeting on March 14, 1975
.
SARB meeting on May. 15, 1975
'
(2) Abnormal Occurrence 23-74 SARB meeting on.May 15, 1975
,
l.
(3) Emergency Plan Review SARB meetings on November 6, 1974, and May 16, 1975'..
'
(4) Unusual occurrence 75-02 SARB meeting on May 16, 1975
- 15 -
,
I
.
+
.
.
.
-
s
-
r (5) = Abnormal'. 0:currence -75-01 SARB meeting ~on March 14, 1975.
s (6) Special Report 21 (AO.75-01).
1.
SARB meeting on.May?7, 1975.
.
(7) Reduced Incore Instruments. Operable
-
,
15ARB meeting on November 6,2 1974.-
'
-
(8) -Increase. of Reactor pover' from 80% to' 90% <(Reduced.
~
,
lIncore Instruments).
,. SARB meeting on-November'22,' 1974
'
^
k.
- The inspector reviewed the Quality., Assurance program as
- established for operations and noted a number of significant
'
itens:
,
.
(1) Plant operations surveillance audit schedules-ha'd been
~
-'
established. The: inspector-reviewed three recently.
.
. completed audits by the QA group..
Emergency _ Diesel Gener' tor. Test.(Augusti7~L1975)~
a
,
Canal L'ater Samples (August 8,1975)
-
Pipe Tunnel. Inspection (August. 12, 1975)
-
No apparent discrepancies were noted on this-checklist.
,
_
L(2) An audit of' operations;had.bcen. completed,. comparing the
- Quality Assurance requirements-with the. implemented' pro--
cedures in the plant. The inspectors noted that the plant
~
superintendent had been informed of-the inadequacies.and-a management-meeting was planned to resolve the deficient areas.
(3) The. inspector noted that the Quality Assurance. program-
is in the.early stages of implecentation and it appears the schedule for fully.i=plementing the QA program vill be co=pleted by approximately October 15, 1975, as. indicated, by Consumers Power Company 1to the Division of Licensing.
3.
Shutdown margin test procedure performed on June 5,1975 (BRP-RE-03,
-
Revision 3).
The inspector reviewed the documented performance of the test procedure and noted'the following:
,
During fuel movements which were performed on February 20, 1975, a.
and February 25, 1975, Attachment A of'the procedure was per-formed as required b'y; Technical Specifications 5.2.2(b).
.
16 -
-
.h s'
!
t
-
.
-
-,,.,.
,,a-
,-
..l,
-,A,,-
,., -.
,,
.. - - -
,
-
e
'
,_ _ _ _. - - _ _ _
_
__
_ _ _ _ _ _ _ -. -
- - - - - - - - -
.---.-- --
___
b.-
Attrchment'B was not ptrfcrmed ct the b2 ginning ef~cycis 13-becaus2 the computer esiculations indient:d.the reteteri would go critical during the tect.-
^
'After. substantial core' depletion occurring'between July of c.
1974, and January. of 1975, it was determined-that the -
Attachment.B could be performed..On June 5, 1975, immediately
-
_
i prior to the return of.the plant-to power operations following'
.
the extended outage for corrective action. associated with A0
-
1-75, the test.was performed.
d.
'The purpose-of-the test was to' determine whether a:shutdowni margin existed with two adjacent control; rods withdrawn. The test was optional and-if successful.would allow continued opera-
'
tion in the-event a single rod.became inoperable in the'with-drawn position.
.
c.
The inspector review of the procedure (BRP-RE-08,' Revision 3)-
and the Reactor Log Book (Book 85,'Page'65) determined that during the evolution, the reactor was critical on a 138 second period at 2030; and at 2038 all rods werc. inserted terminating the test. The inspectors discussedfthe apparent' inability to determine that criticality was imminent in conduct of the-test.
The licensee representative cxplained the.s'ource-to-fuel region-to-detector geometry and the range of notch worths which are-encountered in the test. -The inspectors asked the licensee to consider measures which'would give improved' visibility to
' reactivity increases during future tests or minimize reactivity incremental natch worths during future tests.
4.
RT r cedures and Calculations (BRP-RE 2)
NDT L
..
a.
The inspector reviewed the RTNDT pr cedure perforced in April-1974, by J. F. Ruemin and reviewed by R. W. Voll on April 25, J.J.FremeauankDT.ProcedureperforcedinFebruary-1975,'by 1974; and the RT reviewed by R. Able on February 25, 1975, in-accordance with Technical Specification 4.1.1(i)
No apparent discrepancies were noted.
5.
Estimated itical position prediction error. The. inspectors
-
reviewed with the Reactor Engineer the approach to c-iticality on June 7, 1975, at 0306. The inspectors determined that the Reactor Engineer was aware of the discrepancy between the computer pre-dication and his manual calculations performed for a reactor operator training startup in March of 1975. - The inspectors' determined that the Reactor Engineer had resolved the discrepancies to satisfy himself that no unsafe condition existed, but the discrepancy was not brought to the attention of management to obtain review and
- 17 -
i r
.
..
.
.
etncurr:nes. [ Adminict*etiva proc dura l'.3.D.3 cppests to require Er:portingLc:nditions coverse to quality. rclating to cafety-ralated items and activities.
Criterion XVI'of Appendix B to 10 CFR 50 requires:that " measures to be established'to1 assure that conditions.
adverse to quality;are promptly " identified... corrected, documented'
and reported to the appropriate levels of management."- Section 7
,I of the Big Rock Point Technical Specifications-notes management
~
review of acts,' incidents,:or other' situations which may compromise
.
.
~
the safety. The inspectors noted this item of noncompliance
- ba ~a previous inspection. report, but;the.importance of identification
~
and review of noncomforming items required further discussion with the reactor engineer personally and, plant management.
6.
. Unresolved items from Inspection Report 050-155/75-05.. The contain -
-ment individual component lcak testing (Containment' Ventilation Supply Valve (CV-4090) concerning the_ containment side bolted flange lenk test, the containment' ventilation supply valve (CV-4097).
Icak test, and the emergency escape lock,1cak test; and the design adequacy of-the containment sphere ventilation system were reviewed-by the inspector..
.,,
a.
The containment ventilation supply valve (CV-4097); with the
containment-side bolted flange. seal which was-removed and-replaced on April 10, 1974, was not tested in accordance with the requirementsLof section 111.B.2~and IV.A'of Appendix J-to 10 CFR 50. 'This Type B test requires that the test pressure to be Pa (23 psig) after the valve replacement. - The failure to perform this test will be continued as an unresolved item with respect to Ar,)cndix J to 10 CFR~50.
b.
The containment ventilation supply valve (CV-4097) which was removed and replaced on April 10,_1974, was tested, after
,
,
completion of the. maintenance evolution and retested at six months intervals in accordance with Appendix J_ to 10 CFR 50-as per test procedure T180-1 in the opposite direction tq,~the flow under the design basis accident.
The practice of perform-ing a Type C test onTa similar replacement valve, in a' system which does not allow testing in the direction of the design basis pressure and flow, is acceptabic, The Emergency Escape Lock outer door and equalizing valve has not c.
been tested at six months intervals at the design basis accident pressure of Pa (23 psig). This Type B test at Pa (23 psig) is regiured by sections III.B.2 and III.D.2.
The failure to perform this test will be continued as an unresolved item with respect to Appendix J to 10 CFR 50.
- 18 -
.
.
.
d
.
.
m
.
.
.
.
.
. -
...
.
.
.
d.-
The Type B' testing of tha emergency excxp3 lock sftar each-opening will,.for the interim, not be required since the
. design.is such that the door seals cannot-be readily tested.
The containment ventilation supply ~ system design adequacy-is e.
,
under study by the Division _of Reactor' Licensing.
7.
The inspectors' reviewed the operation of th'e normal ~ personnel access hatch, the equipment. access hatch, and the emergency escape lock.
The personnel access hatch and the equipment access hatch are a.
of similar design and operation. The?rcspective door-locking.
pins which _are' operated -(inserted and retracted) by. rotation of the interlock handles also operate the spring-loaded equal-izing valves. ~The equalizing valves _ provide a pathway to-equalize the hatch internal pressure with the extcanal pressure, allowing the respective door to open freely and safely. 'During the review of.the door operation, including local' observation of.
~
the equalizing vaF._' operations, the inspectors observed that
_
the equalizing va.8vc associated with the-door last utilized
~
appears to remain in the open position. This' occurs because the'
mechanical interlock handles / equalizer valvc_ operator _are left in other than the mid-travel position. The boundary, provided by the associated door and equalizer valve, appear to be breached during normal operation.
In a_ previous item of noncompliance con-.
cerning the emergency escape lock inner door operation with respect-J to 10 CFR 50, the NRC position was stated.J.p2p f Appendix to Technical Specification 3.4.2 and Section 1me routine operation of the personnel access hatch and the equipment access hatch mechanical interlock hand 1cs/ equalizer' valve operator in
this manner is considered contrary to the intent of1the Technical Specification and Appendix J.to 10 CFR 50.
This item will be , carried as an unresolved item pending acticn and review by the licensee.
By-telecon with the licensee on August 25, 1975, the inspector verified that the equalizer valves would be maintained in the closed position except when entering or existing the' access hatches and that the lack of operating procedures for the hatches would be reviewed.
b.
The_ review of the emergency escape lock operation revealed that the equalizing valves are Icf t 'in -the closed position.on the respective door by design of the operating mechanism, when the door and door operating lever are in the closed position.
The inspectors noted during the plant' tour that the emergency access hatch inner door was being maintained in the normally closed position.
21/ Ltr, IE:III to CP, dtd 6/24/75.
22/ 1R 050-155/75-05.
19 - - ( . g.. . .x
8.
Safety. Limits, Limiting' Safety System Szttings, cnd Limiting-Conditiens for~0 par: tion. The inspector rtviewed:salected plant-maintenance items, facility change items, and operating records to assure. plant operations within the license requirements.
a.
Selected Maintenance Evolutions'
, t- - (1) Reactor Steam Drum (RSD)' dated. April 8, 1974, Valve packing.
. (2) LCore Spray System'(PIS) dated May l',~1974, pressure- ' Switch.
(3) Emergency Power System (EPS) dated May 17, 1974, Battery Charger.
(4) Reactor Coolant System'(RCS) dated June 15, 1974, Recirculation Pump Seal.
(5) lbin Steam System (MSS) dated. July ~ 11,1974, Limitorque' Operator.
(6) Emergency.Condenscrl System. (CIS) dated July _15,1974,. Limitorque' Ope.rator.
(7) Containment Isolation System (CIS) dated February 6, 1975, Snap action switches.
(8) Control Rod Drive System (CRD) dated December 16, 1974,' CRD inspection.
(9). Control Rod Drive System (CRD) dated February 17,-1975, Falay replacement
No apparent discrepancies were noted.
b.
Facility Change Activities (1) Control Rod Drive' System (CRD) d'ated April-5,1974, Roller removal.
i' (2) Containment Isolation System (CIS) dated July 8,.1974, . Solenoid. replacement.
(3) Primary Coolant System (PCS) dated April 6,1974, Limitorque gearing replacemen't.
No cpparent discrepancies were noted.
. % so - - . e
, . - c.
Pirnt Systems The inspector verified specific : plant system' conditions and parameters to insure operability.- , (1)- The containment access hatches doors were verified in the closed position with the exception ofJingress and egress.
(2) - The semi-annual containment component' leaktest' performed - during March of 1975 was reviewed to be within Technical Specification'and Appendi of 10 CFR 50 limits except as noted by the licensee .(3) ~As noted in' Report Details, Paragraph 4, of-this report. the ~ latest'RTET calculations were reviewed.
~ ' ~ (4) The operability status of the ettergency'diescl generator during' June and July of 1975 was reviewed.
. (5) As noted in Report Details, Paragraph 3 of this report, the core shutdown margin verification check was. performed on Lbruary 20 and 25, 1975.
(6)' The inspector verified that the control rod sequence was established and was being utilized.- During the inspection the inspectors noted that one rod was inoperable and tagged - (E-4).
The' inspectors verified no other rods tagged as-inoperabic.
(7) The three power range neutron flux channels were noted as operabic.
' - (8) The two startup range neutron monito-ing channels and two intermediate range channels were verified operabic during the shutdown margin tests and plant startup on June 5-8, 1975.
No apparent discrepancies were noted.
9.
The inspector review of the performance of'the conthly surveillance Test T 30-14, Core Spray Heat Exchanger Leak' Rate Test, as a . followup of A0 5-74 and A0 23-74 revealed that the test was;no.
performed on February 24, 1975, March 31, 1975, and about April 24, 1975. The inspector verified that the test was perforced on June 2, 1975, ic=ediately prior r3 plant startup. The inspector noted no documented review of the failure to perform the surveillance 23/ 'A0 050-155/08-75, Ltr, CP to DL, dtd 3/31/75.
- 21 - ' ' . , s
. . tacts, in ccesrdanca with'Adminictrative Preciduro-1.3. 'Th2 inspector verified thatlthe: plant'was in the cold shutdown. con- . dition with -the motor' opera tor on valve NO-7072 removed for testing purposes while performing corrective actions associated with . A0 050-155/01-75.
, .- , ~ 10'. Nacility' Tour , .a.
.The inspectors' observed'that the cleanliness of-plant areas.
were adequate with the only area of concern noted as outside the containment. equipment hatch duc'to the. birds roosting- ~about'the loading ramp.. b.
The inspectors observed that the. radiation areas were localized-and with these restricted areas, the facility was available to ~ normal entering and exiting, within the normal plant security: control' system.
The inspectors noted no equipment leaks'or controlled areas c.
, ' caused by leaking: systems or components.- The inspectors verified that the combustibic materials 24/ d.
- (cardboard boxes) had been removed from room 441 in the containment vessel.
Theinspectorsobserygpthatthecondensate.tankventedtothe c.
outside cnvironment.-- - Even though the tank.contains and inter-nal overflow arrangement, it was not determined whether.the internal overflow could accept at the. full return flou from the - condensate pumps during large load transients.
The licensco's representative. indicated that a review would be performed to: insure that' maximum return-flow could be' handled by the ' internal overflow at the maximum operating icvel in'the tank.
The inspectors noted that the tank-was not insulated, and questioned the licensee if an icing prob 1cm existed, or uas ever encountered, during the winter months. The licensee indicated that up to 18".of ice forms on the inner walls of the tank, but _this does not cause any operating probices.
. f.
The inspector's observed that the operation of the emergency escape lock inner door operating lever l(on the outside of the escape lock)' appeared to be_interferred with by-the conduit to the neuly installed emergency ' escape lock outside security door.
g.
The inspectors observed that the station battery for the facility did not have lateral nor_ vertical. supports to. prevent movement and possibic damage, during a seismic event. cThe 24/ IR 050-155/75-07.- 23/ IR 050-155/74-08.
? - 22 - t l' . e- - -e
a , ~ Finni Hazards Summary' Report Section !2.6. indicates the plant . structural' design complies with the Uniform Building Code . (UBC) and horizontal forces based on Zone 1 are used. The con - tainment vessel, concretc. structure, and the equipment '(safety-1related) within are-designed to approximately twice the UBC ( requirements; or a conservative seismic factor of 0.05.
The station battery provides power for the' safety-related valves at the facility and would appear to require scismic supports,Lwhich' would assure continued service during the design ~ seismic event.
The licensce's_ representative indicated this area would be investigated.
h.
The inspectors noted that the paint on the'inside_of the~ cont'in-- ment sphere wall in the area of the cooling water heat _exchangers was flaking and peeling. The licensce's representative indicated-this arca would be investigated.
1._ The inspectors verified that the controlled area exists above the core spray room (427) and the maintenance shop roll-up ' door wcre more closely monitored. A personnel frisker was mounted at the
exit to the off-loading ramp, and the maintenance shop roll-ups door was being maintained closed.
j.
The inspectors noted.that the floor drain located at the entrance to the enre spray pump /hcat exchanger room (427) was not marked. The drain appeared to be a storm drain, rather than one for radioactive l'iquids. The licensce's representative stated that the dra4 path would be' determined and the drain marked if required. 9j , 11.
The inspector, accompanied by the shift supervisor, observed the coergency.
dicsol generator ucekly surveillance test (T30-13) conducted on Thursday, August. 14, 1975.
The diesel engine started immediately-upon initiation.
No apparent discrepancies'were noted by the inspector.
. ! , 26/ IR 050-155/75-10.
]]/ Ltr, CP to DL, dtd 9/27/74.
- 23 - ! , s I.
_. }}