ML20210M567
| ML20210M567 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 01/12/1987 |
| From: | Clark C, Jim Melfi, Richards S, Wagner W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20210M507 | List: |
| References | |
| 50-312-86-38, GL-85-05, GL-85-20, GL-85-5, IEIN-85-042, IEIN-85-42, IEIN-86-056, IEIN-86-063, IEIN-86-56, IEIN-86-63, NUDOCS 8702120575 | |
| Download: ML20210M567 (26) | |
See also: IR 05000312/1986038
Text
.
+
c p o;
,
~ # --- - --
- g . -
. .,
4
, :
,
4 -
. 3. 2
,
-
'
H g. , 7,- -
.s,; -
,
,
'
i ** - , * r.
> ,
p. -
.
"
-
,1 , sj ;
.
,
-
- a y
'
_
,
- ; to
" "
< 'o -
-
U.'S." NUCLEAR REGULATORY COMMISSION
' '
('
'
. . i. #
.
,. , 'REGI0'N Vl
~
. , ,: .r
"* Report No. 50-312/86-38
-
.
Docket No. 50-312; . . "
'
,
License No. DPR-54
.
,,
.
Licensee: -Sacramento Municipal Utility District '
'
P. O. Box 15830- .
Sacramento, California 95813; ;
~
= Facility:Name: Sacramento Municipal Utility District (SMUD)
- *
-
Inspection Condtieted: November 17-21 and December 8-23, 19867
Inspected by: [M , ' / - 8-- 8 2 -
C. Clark, Reactor Inspector Date Signed
M .
b Y-h
E Melfi, Reactdr Inspector
.
. Date Signed
k?d. // lasani / - 7-B7
Date Signed'
'
W.pner,Reac 'r~faspector . ,
,
, Approved by: S.h . / - /2- 97-
S. Richards, Chief, Engineering Section Date Signed
'
2
Summary: ,
! '
,
Inspection on November 17-21 and December 8-23, 1986 (Report No'.' 50-312/86-38)-
'
Areas Inspected: Routine unannounced inspection by. regional based inspectors
,
of licensee action on inspe'ctor-identified items, Licensee Event. Reports,.
! open items, I.E. Information Notices, Part 21 and generic letters.
- Inspection procedures 30703, 92700,.92701, 92701-1, 92702, and'92703 were
covered during this inspection. '
'
Results: In the areas inspected, one violation was identified for failure to
include appropriate acceptance criteria for snubber' lock-up velocity
- (paragraph 2.k).
- 3
~
.
,
T
4
4
, -
^ ~
4
'
l -, -
-;:
'
-),
,! 8702120575 870122
PDR ADOCK0500g2
G '
> <
'
I ,
'4 - . , , - - .__ __. _ . . - .
. .
DETAILS
1. Personnel Contacted
- D.'Poole, Plant Manager
- B. Croley, Deputy Plant Manager
- G. Coward, Deputy Restart Implementation Manager
- S. Knight, QA Manager
- D. Army, Nuclear. Maintenance Manager
- R. Colombo, Regulatory Compliance Superintendent
- T. Shewski, Quality Engineer *
H. Heckert, Staff Assistant (Acting)
- J. Browing, Regulatory Compliance Engineer
- J. Robertson, Nuclear Licensing Engineer
- Denotes those who attended the exit meetings.
2. Licensee Action on Previously Inspector Identified Items
a. (Closed) Unresolved Item No. 50-312/83-22-02: ' Approval of Proposed
Amendment 97 to the Technical Specifications
Table 4.1-1, " Instrument Surveillance Requirement," in Amendment 54,
of the Rancho Seco Unit 1 Technical Specifications, contained
typographical errors in the test column for items 48.a and b. The
test column for item 48.a should have read NA and item 48.b should
have read M, but instead they were reversed. To correct the above
typographical errors, proposed Amendment 97 tc. the Technical
Specifications required approval.
The proposed Amendment 97 was approved February 21, 1985, and issued
as Amendment 60 to the Technical Specifications on March 8, 1985.
The inspector reviewed items 48.a and b of Table 4.1-1 of Amendment
60, and found the typographical errors have been corrected.
This item is closed.
b. (Closed) Followup Item No. 50-312/84-26-02: Program for Changing
Procedures to Reflect Technical Specification Amendments
The licensee had been previously requested to examine their existing
program for updating operating procedures and to evaluate any
program modifications necessary to ensure that procedures are
implemented in a timely manner to Technical Specification changes.
This evaluation resulted in the recently issued Administrative
Procedure AP.72, " Technical Specifications Amendment Procedure,"
effective date of October 15, 1986. The inspector's review of AP.72
revealed that Section 4.6 requires additional actions of the Plant
Review Committee (PRC) with respect to processing a Proposed
Amendment to Technical Specifications. Essentially,-it requires the
cognizant engineer to present the change to-the PRC. Section 4.6.1
then requires any PRC members, for which their department procedural
changes will be required upon NRC approval of the Proposed
Amendment, to make known any desired issuance delays after NRC
~
ri
, .
approval. LIf no requests for delays are made, then-the NRC_ approval
date and the effective date of the operating procedure will be
' coincident. In order to ensure these procedures are implemented in
a-timely manner, Section 4.6.2 requires PRC members to retain copies !
of the Proposed Amendment and to use the interim.between the
proposal and approval to draft procedural changes for immediate
implementation after amendment issuance. The licensee's evaluation
and subsequent issuance of AP.72' adequately addresses the
inspector's request.
This item is closed.
c. (Closed) Followup Item No. 50-312/84-31-01: Quality Assurance
Review of Bulletin Response System
The licensee did not provide a response to a March 10,.1983,
bulletin until July 30, 1984. The bulletin requested a ninety-day
response. The inspector was concerned about the timeliness of the
licensee's response.
In January 1985, the licensee's Quality' Assurance (QA) department
committed to audit the system controlling bulletin responses and
provide some action to prevent further delinquent responses.
During January 14-17, 1986, the licensee performed audit No. 0-777
of NEP 3104.1, .2 and .3 as they apply to the control of and
response to NRC I&E Bulletins. The summary for this audit report
stated in part, "The procedures used to receive, control and respond
to NRC correspondence are apparently adequate to control the
~
response to I&E Bulletins." The audit reviewed the Coordinated
Commitment Log (CCL) for I&E Bulletin commitments and overdue
commitment responses. No overdue responses were found.
In response to QA Audit 0-777, the licensee stated in a memorandum
NL-B6-127, dated April 11, 1986, "The District has pledged increased
management emphasis on commitment tracking and submittals-to
regulatory agencies....This process will'be considerably streamlined ,
and improved with the development within the next month of a new
commitment tracking system. Licensing has contracted with Stone and
Webster ' Engineering Corporation (SWEC) to develop and implement the -
new system, which will be fully operational by May 1, 1986."
The inspector reviewed the available documents identified above and
noted that the new commitment tracking system had its start
date changed until after September 1, 1986. On December 15, 1986;
the licensee signed out directive ND-86-19-A titled, " Commitment
Management," which implemented a new coordinated commitment
tracking system (CCTS). The new CCTS included all the information
originally contained in the CCL system, and should improve the
licensee bulletin response system.
This item is closed.
.
A
%
=__ _.m. J
-
.-
,. ,
,
,
. ,
' *
3.
.:
.
'
i ,.
'
d. (Closed) Followup Item No. 50-312/85-04-01: -AFW' Start with-MFW
Pressure Signal Testing
The inspector identified tha'tone of three automatic: start signals,
low main feedwater header pressure, was not being tested during the
'
eighteen month shutdown surveillance. .However,'the licensee does
-
use this signal on their monthly auxiliary feedwater pump
surveillance test. ; Therefore, although the pumps :are-not:being-
started with the low main feedwater header pressure signal during
shutdown, per the technical specifications, the-licensee has(shown
operability of-the start signal on a monthly basis. -In order to
clarify the auxiliary feedwater surveillance testing requirements
the licensee submitted to.NRR on: June 13, 1986,. Proposed Amendment-
No. 148. The Proposed Amendment > revises Technical Specification 4.8
to permit system testing of the auxiliary feedwater pump under
conditions of either power operation or plant. shutdown. : Subsequent -
to this amendment request, the licensee:has proposed a change to the
'
' Technical Specifications which'will permit operation of the
Emergency Feedwater Initiation and Control .(EFIC) System. This
Proposed Amendment No. 152, . submitted to NRR on December 5,1986
-
(letter JEW 86-713) will incorporate the auxiliary feedwater tests
requirements of the previously submitted Amendment 1148. The
inspector reviewed the documents submitted to NRR and is satisfied
with the actions taken by the licensee to address this item.
This item is closed.
e. (0 pen) Followup Item No. 50-312/85-04-02: Review and Verification
of Past Commitments and Design Implementation
This item was generated.as a result of a commitment to install
hydrogen monitor vent valves as a NUREG-0737 requirement. Since the
hydrogen monitors penetrate containment, and do not receive a Safety' '
Injection Actuation Signal (SIAS), they were required to be locked
closed. This hardware change was performed by Engineering Change
Notice (ECN) 2938. The Design Basis Report (DBR) found in the' major
portion of this ECN states that the valves shall be administrative 1y
locked closed. The actual work done for_this item is by sub-ECNs as
per procedure NEP 4109 (Rancho Seco Configuration Control
Procedure). 'The part of the commitment that failed was not'the ~
hardware installation, but the administrative controls (a software
item).
The inspector talked with licensee personnel about ensuring that'
~
past commitments were implemented. .The licensee is currently
. writing a procedure to-identify,. address, track and assure
completion of all commitments madelpreviously. The-commitment
evaluation program project procedure is . currently;in a draft form.
This item will be closed when this procedure'is finalized and
inspected for its adequacy in verifying that hardware and software
commitments are completed.
_, , _ _ _ .
,,_
,
L
+ -
4
f. (Closed) Notice of Violation No. 50-312/85-08-01: Battery
Maintenance Procedure and Data Errors
The licensee's. response to this violation was previously reviewed in
Inspection Report 86-25. The item which remained open concerned'the
finding that during the initial inspection (Inspection Report No.
50-312/85-08), Procedure EM.106, Revision 4, did not specify the
step to be used when starting an equalizing charge without-
performing'a discharge first. Licensee electrical maintenance
personnel were using applicable'section of EM.106 to equalize the
battery when required. Also, during this initial inspection, a
review of battery test results identified errors.in the recorded
data. A followup inspection in July of this year found that the
latest issue of Procedure EM.106 had not been revised to address the
weakness identified in NRC Inspection Report No. 50-312/85-08 and
the licensee could not identify what actions were taken in response;
'
to errors identified in battery test results data.
,
~
During this inspection, the licensee agreed with the inspector that
they did not have an existing written procedure, with specific steps
that maintenance personnel should follow to place a battery or
battery cell on equalize. Procedure EM.106 will be' replaced with
EM.106 A, 106 A2, 106 B, 106 B2, 105 C, 106 C2, 106 D, 106~D2, 106-
E, and 106 F, to cover battery testing. The' licensee stated that a
new procedure (EM.151 - Equalize Charging of Batteries) will be
issued prior to restart, for maintenance personnel to follow when
charging a battery or battery cell, if required by battery
surveillance.
In response to the errors identified in the initial review of the
battery test results, the licensee provided a memorandum dated
July 15, 1985, from C. Linkhart to S. Crunk, which stated in part
the following:
(1) " Existing procedures shall be rewritten to upgrade them, to
eliminate a majority of the incorrect data seen on old
procedure data sheets."
(2) " Maintenance Engineers will review all procedure enclosures /
data sheets instead of a maintenance foreman. This not only
allows more time for review, it provides a fresh look at the
data by a completely independent person."
(3) " Rough data from the field will no longer be copied onto a
fresh enclosure. This practice was instituted with the good
intention of providing nice, clean, presentable data for the
history file. Unfortunately, it has too often resulted in
transcription errors."
(4) "The comparison of new data to old data.has not been a formal
process up to now. Our rewritten procedures will provide
formal accounting of this comparison with guidelines for action
to be taken when a negative trend is discovered."
.
.
.
.
.
'
s
5
Based on the inspector's
procedure changes review
surveillance data and maintenanceof new
changes in methodsand proposed proc
this violation were equate.
adthe inspector c ofconc,ludedand
handling
,
th
that the licensee corre
ccommitments
ve tiabove
actions for,
g.
This item is closed. k
(Closed) Followu
Re uired to Delete 50-312/85-27-02: RefereItem No.
AP-27 Revision
b
Action," dated August 10CI-7Revision 1 to Quality A
QCI-7,
issued. " Corrective suranceAction Procedure
" whi h (QAP) No
,1984,
,
c referenced Quality Contr l27, " Corre
o
The inspector
1986.
had been cancelled ever and n Procedur
This
Action." e:, revision
"Procedur deleted step 7 ireviewed Revision 2 to
" which referencednthe the paragraph titledo. 27 dated Jan
,
original QCI-7 titled "Co
h.
This item is closed . rrective ~
(0 en) Followu
Em sProcedures to Assure PrItem No.
o 50-312/86-07-07: .
",
Inspectio er Control of NoncondensiblLicensee e Gases in an
to Re ..
followup, n Report
50-312/86-07
operators which required examinatiidentifi
ed an
gases whenever the pressu
r zer a eiare
on of aware of actions
to procedures to assuropento t k item 3 for
The inspector empties. e that h\
control noncondensible /
p (1) Operatin reviewed the following do 4
.i
cuments:
-
j { t
System,"g Procedure (0P), A.74, Revisi -
(2) dated June 4, 1986c - Se ti
OP A-1, Revision 20 on 20, " Control Rod Driv
on 3.5. e
1 {1
September 5,1986
,
(3)
,Sec ti" Reactor Coolant Syste
on 3.22. m," dated
thI
t OP B.4, Revision 40
y September 5, 1986 ,Se ti" Plant Shutdown
c U
The on 3.28. n !
and Cooldown," dated
event re
above documents had the f l
- '
tripped) quiring RCS venting occurso lowing instructions add d
to the
Therefore,
prevent operation
the in limit (see
inspector , all CRDs shall be e , "If an
run in (not
,
licensee's
of gas bound CRDs
emergency procedure. er had
.
measures to
been icould <inot
$,
l
\
t
This item remains
open pending further
mplemented in the
~
review. $]m .
17,
-
j
}
\
.
.
-
,- -
,
. .
5
Based on the inspector's review of new and proposed' procedures,
,
procedure changes and maintenance changes in methods of handling
surveillance data, and the licensee's specific commitments above,
_
the inspector concluded that the licensee corrective actions for
this violation were adequate.
This item is closed.
g. -(Closed) Followup Item No. 50-312/85-27-02: QAP-27 Revision
Required to Delete Reference to a Voided QCI-7
Revision 1 to Quality Assurance Procedure (QAP) No. 27' " Corrective-
,
Action," dated August 10, 1984, referenced Quality Control Procedure
QCI-7, " Corrective Action," which had been cancelled and never
issued. '
.
The inspector reviewed Revision 2 to QAP No. 27 dated January 1,~
'
1986. This revision deleted step 7 in the paragraph titled '
" Procedure:," which referenced the original.QCI-7 titled." Corrective,
Action." ,
This item is closed. - >
h. (0 pen) Follow 9 Item No. 50-312/86-07-07: Licensee'to Reexamine
Procedures to Assure Proper Control of Noncondensible Gases in an
Emergency
Inspection Report 50-312/86-07 identified ao_open' item 3 for-
followup, which required examination of procedures to assure that
operators are aware of actions to take to control noncondensible
gases whenever the pressurizer empties.
The inspector reviewed the following documents:
(1)' Operating Procedure (OP), A.74, Revision 20, " Control Rod Drive
System," dated June 4, 1986 - Section 3.5.
(2) OP A-1, Revision 20, " Reactor Coolant System," dated
September 5, 1986 - Section 3.22.
(3) OP B.4, Revision 40, " Plant Shutdown and Cooldown," dated
September 5,1986 .Section 3.28.
The above documents had the following instructions added, "If an
event requiring RCS venting occurs, all CRDs shall be run in (not
tripped) to the in limit (see...for venting requirements)."
Therefore, the inspector could not determine whether measures to
prevent operation of gas bound CRDs had been implemented in t' e
licensee's emergency procedure.
This item remains open pending further review.
_
__ _
_. _ _
t
. r ,
.
S
,
i. (Closed) Notice of Violation No. 50-312/86-08-02: No Control of '
Measuring and Test Equipment (M&TE)
,
The licensee's Administrative Procedure AP12 (Plant Housekeeping ~and
Inspection) required that " Tools and test equipment shall be stored
in their proper location'at the;end of the workday and anytime when
not in ose." <
Contrary to these requirements, an inspector found, on two
occasions, calibrated tools uncontrolled at the= work site, when no
work was being accomplished. On March 5, 1986, and again on
March 7,1986, an inspector observed unattended calibrated tools
placed on a tool cart, in cardboard boxes, and on the floor of the
computer room adjacent to the control room.
In response to this item, the licensee had each piece of reported
equipment checked to ensure that inadvertent damage had not
occurred. Each item was subsequently found to be in proper working
order. Additionally, to ensure that greater care will be exercised
over calibrated equipment in the future, the electrical maintenance
superintendent issued verbal instructions to place M&TE within
carts, tool boxes or cabinets while not physically in use. These
verbal instructions-have also been included in Revision 5 (issued
June 30, 1986) to Administrative Procedure AP.33 (Calibration and
Control of M&TE), in paragraph 6.3.2 which states ". . . All
individuals and supervisors must not allow M&TE to be left in any
area when it is not being used and must ensure that M&TE is returned
to the appropriate storage area after use."
The inspector reviewed the applicable licensee documents
(Administrative Procedures, responses, etc.) and it appears the
licensee has taken the necessary' corrective action to prevent
recurrence of this item.
This item is closed.
j. (Closed) Notice of Deviation No. 50-312/86-18-08: Failure to
Satisfy Commitment
This deviation addresses the licensee's failure to satisfy *a.
commitment to make a permanent revision to Procedure I.103 by -
February 28, 1986. The licensee responded to the Notice of
Deviation in a letter (JEW 86-223) to Region V dated July 21, 1986.
The response pointed out that there has not been a need to
physically perform the power range nuclear instrumentation
calibration since the plant was' shutdown on December 16, 1985. In
addition, the Power Range CalibrationfP rocedure, I.103, can only be
_
done when the reactor is generating enough power to be measured by
the ex-core detectors. The inspector verified that the licensee had
, revised Procedure I.103 as they previously committed to do. Also,
the inspector reviewed a draft Management Directive which, when
approved, will apply to the identification, tracking, implementation
,
and closure of commitments by the District to regulatory and other
i
l
l
i
k-
--
.-
-
,
.. ,
7
external agencies.' This Management Directive appears to provide the
necessary instructions to ensure timely completion of commitments.
This item is closed.
k. Unresolved Item No. 50-312/86-21-02: Licensee Acceptance of Snubber
Test After First Test Failed
Snubber No. 129 successfully passed its surveillance test after
having failed the same test the previous day, June 26, 1986. The
inspector expressed concerns regarding justification for: declaring
the snubber operable, and why an NCR was not generated when the
snubber failed to meet the acceptance criteria when first tested.
The inspector reviewed QA Procedure No. 26 and verified-that the
procedure was revised to include the requirement that an NCR be
written when surveillancejtest results are not in conformance to
acceptance criteria.
In regards to the operability of the snubber, the inspector has
determined that at the time these concerns were identified,.the
licensee was utilizing a procedure which contained an inappropriate
t
acceptance criteria. Specifically, the acceptance criteria of
Procedure SP 201.10B did not compensate for the effects of
temperature when performing snubber functional tests. Also, at this
time information on temperature compensation requirements was
available in vendor manuals located in the licensee's Technical
Manual Library. A calculation performed by the inspector revealed
that Snubber No.129 fails to meet the acceptance criteria when the
effects of temperature are taken into consideration. That.is,'the
snubber lock-up velocity at test temperature of 78*F was 20 inches
per minute (ipm) whereas the lock-up acceptance limits are between 1
and 18 ipm. Failure of the licensee to include, appropriate
acceptance criteria in their procedure for functional testing of
snubbers is an apparent violation (50-312/86-21-02). ,
1. (Closed) Unresolved Item No. 50-312/86-21-08: Decay. Heat Removal j
(DHR) System Put Into Service Without Initiating Operation of the
Nuclear Service Raw Water (NSRW) System
As an example of a lack of attention to detail in the performance of
routine personnel activities, a train of the DHR system was put into
service without initiating operation of. the NSRW system as required
by plant procedures. This error was identified to the NRC by
licensee personnel after they discovered it.
~
In memorandum NL 86-936 dated December 8, 1986, from H. Sims to R.
A. Little (Subject "CCL #R8608180056, clarification of response"),
the licensee identified what they had considered as the cause of
this occurrence. Operation Procedure (OP) A.8 (Decay Heat System)
stated in paragraph 4.3 (DH removal during RC system cooldown) the
following:
J
.
. '
8
" Initial Conditions
.1 The primary system temperature is <290 F, and~one or two
RC pumps running preferably in Loop B.
.2 The Nuclear' Service Cooling Water System in service to the
DH Cooler as per OP A.24.
.2 The Nuclear Service Raw Water System in service to the
Nuclear Service Water Coolers as per OP A.25."
In order to meet these " Initial Conditions," the Operator should .
follow OP A.24 and start NSCW Pump P482A (P482B), then follow OP
A.25 and start NSRW Pump P472A (P472B). .It was felt that these
steps should ensure the proper operational mode of the-system before
actually beginning the " Procedure" steps of OP A.8.
Obviously, however, there was a case when this did not occur and-
resulted in the noted problem. Therefore,:to eliminate future
recurrences the licensee issued Revision 29 to OP A.8 which added
the following first step to the procedure.
"4.3.8 Start NSRW Pump P472A (P472B) and NSCW Pump P482A (P482B)
and verify proper operation."
While the licensee considered paragraph 4.3.8 redundant, it did
provide a " double-check" to ensure both available water sources were
in service before proceeding with decay heat removal steps.
The inspector reviewed the applicable licensee documents (operating
procedures, responses, etc.) and it appears the= licensee has taken
action to prevent a recurrence of this item.
This item is closed.
3. Licensee Action on Licensee Events Reports (LER)
a. (Closed) LER 83-24, Revision 0: 'B' Nuclear Service Raw Water
Pump Tripped Due to Cable Grounding
This LER reported the licensee's actions in response to a ground
fault that occurred May 19, 1983, in the B phase of the breaker
supplying the nuclear service raw water pump. The B phase cable-was
repulled and spliced to eliminate the ground as part of the initial
corrective action. The other two unaffected phases were also
identified to be repulled at a later date as a precautionary measure
to ensure no additional problems would be encountered with the pump.
The inspector reviewed the applicable licensee documents and noted
ECN (ECN) No. A 4905 was prepared December 13, 1983, to replace the
existing spliced cable with a new 3-I/C 250 MCM, SKV cable.
According to work request (WR) No. 92879 issued November 27, 1984,
the work required to accomplish ECN No. A 4905 was completed May'23,
1985, and the ECN-was signed off completed on June 1, 1985. It
9
'
9.
appears the licensee has completed the original precautionary.
corrective action identified to ensure no additional problems would
be encountered with the pump.
This item is closed.
4
b. (Closed) LER 84-11, Revision 0: Incorrect Configuration Tables in ,
Surveillance Test Procedures
On March 6,(1984, the licensee identified that the configuration
tables in Surveillance Procedures-SP 203.02 A, B,-and C (Quarterly I
and Annual Inspections and Surveillance Tests for_ Hi>I Loop A, HPI
Loop B, and Makeup System Pump and, Valve) were misleading and
incorrect, with respect to the cross-tie isolation valves. The
configuration tables allowed three (3) differentL configurations, one
of which was contrary to the Technical Specifications,;but had never
been used. The configuration tables are allowed to-be used as.
directed by the shift supervisor, but they are primarily used for-
"
information purposes.
As a corrective measure the licensee stated they wouldl revise
Procedures SP 203.02 A, B and C to delete the configuration tables
and reference Operating Procedure A.15 (Makeup, Purification and
Letdown System) for the allowable breaker and valve configurations
for the makeup and high pressure injection pumps. The inspector
reviewed the applicable procedures and verified.they had been
revised as required.
This item is closed.- !
c. (Closed) LER 84-24, Revision 0: Simultaneous Plant Heatup and
Deboration Violated Procedural Control of Reactivity Addition
On November 7,1984, plant heatup was commenced during reactor
coolant system (RCS) deboration. The RCS deboration resulted in 10
ppm reduction in boron concentration over a period of one ~ hour and
26 minutes, while heating up the RCS to 440 F. The core reactivity
at the end of the event was -4.8% Delta K/K, which is 3.8% Delta K/K
more negative than the required 1% Delta K/K shutdown margin. The
event commenced on November 7, 1984 at 1750 when the swing shift
stopped plant heatup in order to perform Surveillance Procedure SP
203.11 (Decay Heat / Core Flood Systems Stop Check / Check Valve Seat
Integrity Surveillance Test) for the core flood tank check valves,
which requires RCS temperature and pressure to be stabilized.
During this pause in the plant heatup, a deboration was commenced at
1835. At approximately 2330, the swing shift was relieved.
_
However, during shift turnover, the significance'of the plant
deboration in progress was not emphasized-to the oncoming shift. On
November 8, 1984 at 0021, the relieving shift supervisor, unaware of
the deboration in progress, directed the control room operators to
start heating up the RCS and then went into the shift supervisor's
office to complete administrative paperwork. At approximately 0226
the shift supervisor noted that deboration was being performed
simultaneously with plant heatup and secured deboration.
'
,
" ' '
10'
,
As corrective actions, thetlicensee performed the following:
(1) Issued a memorandum to the operators emphasizing the importance
and necessity 1.for proper transfer. of information during shift -
turnover.
(2) Conducted a. review of-the shift turnover practices,.which
included discussions with INPO representatives, shift
supervisors and many operators hired from other utilities. A
number of ideas were brought up and ' incorporated in-licensee
procedures. ~The inspector reviewed AP.23 (Revision 20), other
applicable documents and the following memorandums:
i .,;
(a)~ D. Comstock to licensed operators,= dated December 10,
'
1984, on" shift-turnovers;x
(b) D.~Comstock'to G. Coward','dated Februa'ry-13, 1985, on LER s
84-24, CCL 85-0004;
~
^
(c) B. Spencer to shift supervisors, dated March.28, 1985 (S0
5-85),on changes'to AP.23. Also: identified as Special
Order.5-85;
(d) G. Coward to.B.. Spence'r, dated' April 11, 1985, on D.
Comstock's memo, dated February 13,~ 1985; and
(e) G. C. Wallace to distribution,' dated May 9, 1986 (NOS
' _
86-147) 'on Revision 20 of AP.23.
(3) Revised AP.23 in Revision 17'to add new relief / turnover
~
checklists for shift supervisors to power plant helpers, and
two equipment checklists for a control room operator to fill
out during the shift. These documents were added to aid in a
more complete shift turnover and increased communications
between crews.
Based on the information reviewed, it appears that the licensee has
taken applicable steps to ensure a detailed shift turnover, which
should preclude a recurrence of this event.
This item is closed.
d. (Closed) LER 84-25, Revision 0: Reactor Trip
The inspector investigated the LER to ascertain whether the
licensee's review, corrective action, reporting of the event and
associated conditions were adequate.
This LER was generated when the reactor tripped on high pressure due
to a Main Feedwater (HFW)-transient. The reactor trip occurred on
November 18, 1984, during a power escalation. The feedwater
transient was induced by the Integrated Control System (ICS)
attempting to keep up with rapid steam header pressure swings.
Following the trip, the large auxiliary boiler had trouble staying
f
'
11
on line to feed the Auxilary Feedwater Pump. Turbine (AFPT), which
was needed due to steam loads exceeding core decay heat production.
Also, there was some difficulty with the "B" Auxiliary Feedwater
Pump P-318 (turbine driven) steam admission valve (FV-30801), which
had stuck in midposition. In addition, the pump was secured at a
pressure which should have reset the auto-start pressure switch
(PSL-31758) but failed to do so.
The inspector reviewed the trip report and corrective action taken
by the licensee. The root cause of the trip was failure of a
control room operator to keep the governor valve limiter higher than
actual valve position demanded by the ICS, and then rapidly raising
the valve limiter higher which induced the transient. The
corrective action taken by the licensee was procedural cautions in
procedures A.46 (Main Turbine System), B.2 (Plant Heatup and
Startup), and B.3 (Normal Operations) to keep the valve limiter at
100%. These cautions imply that the ICS will now have control over
the governor valves to the turbine under most of the plant
conditions. The inspector was also informed that the MFW
controllers have been recalibrated and now respond faster and more
accurately to changing flow conditions. The steam admission valve
has been added to the preventive maintenance _(PM) program. The auto
start pressure switch will no longer be used, since the AFPT will be
controlled by the EFIC system when it is installed. The licensee is
also doing work on the boilers to improite their reliability.
The corrective actions taken by the licensee should lessen the
likelihood for a reactor trip from the same cause.
This item is closed.
e. (Closed) LER E5-01, Revision 1: H, Monitor System Containment
Isolation Valves Found Open for 7 Days
This LER was generated when the licensee discovered that four
hydrogen monitor system containment isolation valves were apparently
lef t open for seven days. 'This installation was to meet the
requirements of NUREG-0737, item II.F.1, attachment 6. The purpose
of this NUREG-0737 item was to provide continuous indication of the
hydrogen concentration in the containment atmosphere to the control
room. The work on the valves was performed under ECN-2938.
This LER has been addressed previously in inspection report 85-04.
This report left this LER open and also generated three additional
followup items and referenced another item from a previous
inspection report as being similar in nature. These followup items
have been addressed in other inspection reports as follows:
Followup Item Status Inspection Reports
84-19-05 Closed 85-04, 86-36
85-04-02 Open None
85-04-03 Closed 86-18
85-04-05 Closed 85-30
_
' '
12
The one remaining followup item (85-04-02), was for the licensee to
ensure that past design commitments had been implemented. -This item
is described in the followup section of this report.
The corrective actions taken by the licensee include: adding the-
four valves to the locked valve list, the addition of ~ a plexiglass
cover over the controls for the valves, and addition of valve
positions to the IDADS computer in the control room. These
corrective actions were inspected by the inspector and found to be
acceptable. The remaining followup item (85-04-02) will be tracked
under that item.
This item is closed.
f. (Closed) LER 85-03, Revision 0: Incorrect Boron Concentration
Technical Specification Limit
On February 5, 1985, the licensee identified that the Technical
Specification limit for boron concentration during reactor vessel
head removal and fuel loading / unloading was incorrect. This
discrepancy was the result of the fuel supplier basing the cycle 6
'
refueling boron concentration on 'a Keff of .99 rather'than the
specified Keff of .95, and this resulted in a Technical
Specification limit of 1850 ppm versus the correct value of 1936
ppm. This condition had existed since the beginning of fuel cycle
6, on June 17,-1983. The refueling Keff was changed from 0.99 to
0.95 in the cycle 4 amendment to the Technical Specifications; the
corresponding refueling boron concentration was calculated properly
at that time by the fuel supplier (Babcock and Wilcox). For cycle
5, the fuel supplier engineers erroneously used outdated and
uncontrolled Rancho Seco Technical Specifications and reverted to
basing the refueling boron concentration on a Keff of 0.99.
As a corrective measure to preclude further occurrences, the
licensee required the fuel supplier to provide tighter control over
its reload calculations by destroying all noncontrolled supplier
copies of Rancho Seco Technical Specifications. Additionally, the
licensee required the fuel vendor to audit the vendor's control of
Rancho Seco Technical Specifications.
The inspector reviewed the results of Rancho Seco audit reports,
audit No. 0-725 (June 10-13,1985) and No. 0878 (October 22-24,
1986) which found that no uncontrolled /out of date copies of the
Technical Specifications were in the possession of Babcock and
Wilcox. Based on the' inspector's review, it appears that the
licensee has taken'the applicable steps to prevent a recurrence of a-
similar problem.
This item is closed.
<
' - - - -
-
. ...
' '
. J[
- *
-
13 ,-
r
- ,
-g. (Closed) LER 85-04, Revision 0: Fire Dampers not Installed as
Required by Fire Hazards Analysis
On February 11, 1985, the licensee identified that several fire
dampers which were included in their August l', 1977 Fire Hazards
Analysis (FHA) submittal to the NRC, had not been installed.
Amendment 19 to the licensee facility operating license
_
(February 28, 1978) was written based on this analysis. The
-implementation date for the fire dampers of concern was the end of
the 1979 refueling outage. Thus, the licensee failed to implement,
these provisions of Amendment 19. ,
The' licensee stated that: "Previously, the areas for which the fire
dampers were'not installed had been designated,.for-other reasons,
as fire watch areas requiring hourly surveillance; therefore, no'
immediate corrective action was required."
The licensee root cause analysis (incident No. 85-11) dated
September 17, 1985, revealed the following information:
'
(1) The lack of specific details in the'1977 FHA made tl$e
determination and monitoring of the commitment difficult.
(2) The lack of an integrated, district-wide commitment tracking
program did not provide sufficient commitment visibility to the
personnel involved.
(3) The ongoing evaluation of the district's fire protection
program has eliminated the need for some of the originally -
required dampers and has resulted in a revised fire hazards
analysis.
This analysis determined that the root cause was "the lack of
detailed engineering procedures to ensure commitments are properly
implemented."
The licensee corrective actions identified to address the fire
protection concerns were:
Provisions for making the FHA a "living" document undergoing
periodic review and updates.
The improvement of design control, by including a cognizant
fire protection engineer in the review cycle.
Installation of fire dampers consistent with the FHA.
Development of detailed engineering procedures to ensure
commitments are properly implemented.
Additionally, the district developed an integrated,
computerized CCL system to facilitate the logging and tracking
of commitments. This system is now part of the new
_ i
.- -
' '
14 ,
-
b
coordinated commitment tracking system (CCTS)' issued
December 15, 1986.
~
t'
.t*
y
_ ,
The inspector reviewed the following documents:-
- QA Surveillance Activities Reports Nos. 734, 735, 736, 737
(dated October 14, 1986) and 742 (dated October 21, 1986),which
provided feedback ~that the licensee had satisfied the intent of
the identified cbrrective actionsJto be taken'to address the 1 .
fire protection concerns.
- Applicable sections of ECNs Nos'. A-5514, A-5529, A-5767,
R-0763, and R-0764 which were issued.to ensure the installation
of fire dampers was consistent with the FHA.
Based on information reviewed by' the insp'ector, it appears that the
licensee has completed the appropriate actions'to address the fire
protection concerns in this LER.
.This item is closed.
h. (Closed) LER 85-05, Revision 0: Closed Boration Path During Fuel
Movement
On April 20, 1985, the licensee identified that manual valve BWS-041
was closed while fuel movement was in progress. BWS-041 is located
in the flow path between the concentrated boric acid storage tank
and the decay heat removal pumps, which provides the only method of
borating the system in the event of a boron dilution accident during
fuel shuffle. This condition existed for approximately one (1) hour
before being detected and corrected. The improper valve positioning
resulted from a procedure error in refueling procedure B.8
(Refueling Equipment Checks and Core Component Handling) which
opened BWS-041 in step 4.20.15 and then mistakingly closed it while
performing the containment isolation valve line-up in step 4.21.
As a corrective measure to ensure valve BWS-041 is open during fuel
movement, enclosure 7.1. to Procedure B.8 (which is referenced in
step 4.20.15) has been changed per Revision 17 to place a clearance
control tag on valve BWS-041 after it is open. Step .4 of enclosure
7.1 (system check list) reads: " Concentrated boric acid system is
lined up in accordance with OP A.12 with the exception that BWS-041,
' boric acid supply to decay heat' is open and under clearance to
remain open during refueling operations." Once the clearance tag is
installed on valve BWS-041, its position cannot be changed without
obtaining the correct signatures. ' Based on information reviewed by
the inspector, it appears that the licensee has taken responsible
steps to prevent closure of the valve during refueling operations.
,
This item is closed.
,
t
e
, _ . .
7
' '
15 -
.
i. (0 pen) LER 85-20, Revision 0: Essential INAC Flow Controller
Design Error Prevents Auto Control After Return of Power
On October 7, 1985, the licensee iden'ified that the flow
~
t
controllers for the control room / technical support center essential
hVAC. filtration units (trains A and B) were notf functioning in
accordance with the system's design basis report. Specifically,
upon being reenergized following a loss of external power, the flow
controllers would assume the manual mode of operation rather than
the automatic mode. In the manual mede, the flow controllers would
not respond to signals provided to control the air flow' rate within
the Technical Specification. limits. The controllers for both trains
are located in an enclosed box on the roof'of the auxiliary
building. To switch from the manual to the automatic mode in the
reported configuration would_ require that personnel be dispatched to
the roof, remove the enclosure cover and depress the " auto" push
button.
The controller discrepancy was detected while' technicians were
performing an investigation to determine why Surveillance Procedure
SP 211.01A (CR/TSC Emergency Ventilation Systcm Loop "A" Monthly
Surveillance Test) failed.
The discrepancy was believed to be a result of the licensee's
failure during procurement, to explicity state the requirement for
the controller (s) to assume the automatic mode following
reenergization from a loss of external power. The licensee's
Incident Analysis Group (IAG) will perform a root cause analysis of
this event. If the conclusions reached by the IAG differ from the
conclusions of this report, the licensee will submit a supplemental
report.
The flow controllers of concern are the Foxboro 2AC type. To
correct the discrepancy ard meet the system design criteria,
instrument technicians added a jumper wire. to each controller to
ensure that the controller remains in the automatic mode except when
the manual push button is held ia a depressed condition. The
correction was made with the concurrence of the controller
manufac.turer and will not compromise the Class I qualification of
the system. The corrective action was completed on October 10,
1985, through ECM R-0174.
The inspector reviewed applicable licensee documents and noted the
following information:
(1) Rancho Seco Unit I Technical Specification 4.10 (Control
Room / Technical Support Centar Emergency Filtering System)
states: "During an SFAS and a loss of offsite power, the "B"
train of essential HVAC equipment is sequenced to automatically
start upon its actuation signals approximately 6 minutes after
the diesel generator breaker closes."
(2) The Foxboro Instructions MI 250-120 dated October 1984 state in
a note on page 2 under Functional Description: "When power is-
,
' '
16
first applied, the station is forced into the manual-mode until
another mode is selected by-the operator."2
(3) Memorandum from S. Crunk to MRT,. dated September 19, 1986,.
which stated: " Presently the IAG has three' individual.LERs
relating to the CR/TSC essential HVAC (85-20, 84-13 and 86-07)
pending evaluation. Due to the pervasive'and diverse ~ nature of
the HVAC problems, I propose that a combined investigation of
-
the overall problem and causes be performed under one Root
Cause Investigation (RCI). This investigation will follow the
corrective action program and will be completed after the
system has been 're-accepted' af ter successful testing." 7A
handwritten note added to bottom of the memorandum stated:
" Approved at Management Review Team (MRT) meeting dated
10/1/86, will be officially signed at MRT meeting dated
11/4/86."
Af ter reviewing the available information, the inspector identified
the following concerns / questions listed below to the licensee:
It appears that the subject flow controllers were procured with
inadequate purchase specifications. The inspector questioned
whether any other similar controllers / equipment purchased
without specification of mode of operation after reenergization
from a loss of external power. According to a licensee
representative, these two flow controllers were the only two of
that model purchased for this site.
The inspector questioned whether the installation of these two
flow controllers passed the original system acceptance testing,
or any other previous surveillance testing, whether the test
procedures were inadequate or not correctly followed. A
licensee representative stated that these questions have been
passed on to the Incident Analysis Group (IAG) and will be
addressed in the final rcot cause investigation report.
The inspector questioned why over a year has passed since the
occurrence of this Licensee Event Report of October 8,1985,
and the licensee still has not identified a root cause or
addressed corrective action to prevent a reoccurrence of a
similar problem. Licensee representatives could not provide an
answer to this at this time, other than it was a low priority
item. Since the licensee has purchased a large amount of new
,
'
equipment for maintenance, modification and repair work in this
last year, this appears to be an example of insufficient
corrective action. The licensee management apparently should
have taken more prompt action to determine how this happened
and what could be done to prevent a similar occurrence.
This item remains open pending a more complete licensee response.
,
-_-,
7
~ '
<
,
^
' '
17 y
.- ,
j. (Closed) LER 85-21', Revision 0: Emergency Diesel Generator-(EDG)
Auto Start Due to Personnel Error While Trouble Shooting
On November 19, 1985, while the plant'was operating at 83% power,
Emergency Diesel Generator (EDG) "A".was automatically started when
the 4A bus normal supply breaker tripped on an overvoltage ,1
condition. The EDG "A" output breaker closed to the 4A bus and the
bus reloaded as designed; however, the EDG "A" supply fan tripped
shortly thereafter.
~
An investigation of the event revealed that the overvoltage
condition was created when electrical: technicians, who were
replacing a relay in the 4A bus voltage protection circuitry,.
improperly disconnected ground wiring to the circuit's active
relays. Lifting of the ground wiring caused the relays to " timeout"
(trip) on a loss of AC power, thereby resulting in.the bus normal
supply breaker tripping on two-out-of-three overvoltage logic. The
subsequent EDG "A" fan failure resulted from a fan breaker overload
device setpoint being out of the specified range.
To prevent a recurrence of this event, the licensee took the
following corrective actions:
(1). The Incident Analysis Group (IAG) prepared Lessons Learned
Report No. 85-01, which alerts plant maintenance personnel of-
the actions leading to the event and outlined.the proper
precautions to take. The electrical' technicians used
elementary diagrams for this work, which did not reflect the
detail of the connection diagram.
(2) A monthly check of the 3A17 breaker phase overload devices was
implemented for the next 6 months to ensure that the 3A17
overload devices remain set within the specified range and to
determine if drift of the devices was a genuine concern.
The inspector reviewed Lessons Learned Report No. 85-01 and other
available documents. It appears the licensee has taken the
applicable steps to prevent a recurrence of a similar problem.
This item is closed.
k. (Closed) LER 85-24, Revision 0: Shutdown Due to Pressurizer Liquid
Sample Isolation Valve Leak-
On December 22, 1985, reactor coolant system leakage was calculated
by a Control Room operator from frequent dumping of the 120 gallon
reactor building accumulator tank. The leak rate was determined to
be between 15 and 20 gallons per minute, shortly after initial
detection. Unit shutdown action was initiated in accordance with
Technical Specification 3.1.6 which requires the reactor to be
shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of detection of a ~ reactor coolant ' leakage
rate exceeding 10 gpm. During operator actions to identify and
isolate the leak, an attempt was made to close "B" letdown cooler
outlet isolation valve (HV-22008). The valve did not fully close; m
-
=
.,, .
-
, .. -- ,
' ~
-
, 2 ~; .
-
-
18 ,
>
_,' . ,
.
however, it was later determined to not be"in the leak path. The
plant'was brought .to a hot' shutdown condition. The pressurizer
~
liquid sample isolation valve'(SFV-70001).was' closed a'nd the leakage
,
was isolated. . -
-
An investigation of the event determined that the leakage originated
from the packing gland of SFV-70001*. SFV-7.0001 is the inside
containment isolation valve and;is normally closed during power
operation. .It had been opened approximately four (4) hours prior to
the event to allow testing of the Post-Accident Sampling System
(PASS). The valve packing gland was disassembled and the stem
inspected for damage. No damage was observed. Twelve (12) rings of ,
packing were added and the packing gland was adjusted to eliminate
leakage. This corrective action was completed on December 24,:1985.
Following the event, HV-22008 was' examined and found to operate
properly. The valve was stroked from its motor control center
(MCC), the position indications checked, and also positioned and
timed from the. control room. It was further determined that the
valve is not designed to close against system pressure; however, the
valve could be closed under the condition of the event, if
necessary, by first closing an upstream valve such as SFV-22006,
-(letdown to cooler E-220).
The inspector reviewed WR No. 108882, Casualty Procedure C.19
(Letdown Cooler Coil Failure), Revision 8, and other applicable
licensee documents on this subject. It appears the licensee has
taken corrective actions to ensure valve SFV-7001 operation is
acceptable and that Casualty Procedure C.19 has been changed to
first close an upstream valve prior to closing either letdown cooler
outlet isolation valve (such as HV-22007 or HV-22008).
This item is closed.
1. (Closed) LER 86-04, Revision 0: Missed Fire Watch
On March 14, 1986, the licensee determined that three fire doors
were inoperable and fire watches had not been posted within I hour
in accordance with Technical Specification 3.14.6.2. The results of
the refueling interval fire barriers Surveillance Procedure (SP
201.3L) were reviewed, and it was determined that these doors were
~
found to be inoperable. If a fire door is found to be inoperable,
the Nuclear Operations Fire Protection Coordinator should be
informed by the surveillance procedure, and a fire watch initiated
in accordance with an Administrative Procedure-(AP.60).
The root cause of-this item was a personnel error. The individuals-
involved.in this event have been counseled'and retrained to adhere
to procedures.
This item is closed.
-
I
r i
e
_m. ._
. .
9
4. Licensee Action on I.E. Information Notices
'
a. (Closed)'I.E. Information-Notice No. 85-42: Loose Phosphor in
Panasonic 800 Series Badge Thermoluminescent Dcsimeter (TLD) ,
Elements ,
This information notice alerted NRC licensees to a problem noted in
some Panasonic 800 series TLD badges ~that has caused spurious high
readings in one of the badges' TLD elements.
The Panasonic 800 series,TLD badge contains a card that holds four
TLD elements. Each TLD element consists of a thin film of TL'
phosphor attached to a disk backing with a clear teflon bubble
cover. During reading, the phosphor is heated by converging
infrared light on the backing. The luminescence from the phosphor
(which is proportional to the dose received) radiates through the
teflon cover and is read with a photomultiplier' tube.
Several Panasonic TLD users have identified badges where crystals of
the phosphor have detached themselves from the backing of the
element, resulting in high erratic readings in that element. When
viewed through a stereoscopic microscope, phosphor crystals can be
observed sticking to the teflon cover (presumably by electrostatic-
charge). In this position, the loose TL material is not in contact
with the backing and does not get heated when the badge is read.
These TL crystals remain at an elevated energy state and continue to
accumulate dose. Apparently erratically high readings result when
the loose crystals are shaken back onto the backing surface during a
subsequent reading. They are then heated and luminescence
proportional to the total doses received during several read cycles.
This process can cause the affected element to erroneously read as
much as an order of magnitude higher than the other' elements in the
same card. Although the frequency of occurrence is small (one
licensee found only one problem badge in 30,000), there is evidence
that the frequency increases substantially once the badges have been
through 100-200- read cycles.
The licensee stated in a memorandum JR 85-82, dated June 21, 1985,
that at their facility TLD badges were not used routinely for
personnel use but were normally used for environmental trending.
The licensee had not seen any abnormally high or low readings that
would indicate loose TL phosphor. Additionally, the TLD #/ badges
used at their facility had been purchased within the last two to
three years and had not been through 100-200 readings, so they would
not expect any loose paosphor.
The licensee issued a revision 4 (dated December 12, 1986) to
Administrative Procedure AP.308-8 (Panasonic TLD reader) which
referenced this notice in paragraph 2.7 and added information to
paragraph 3.10 for inspection for loose TL crystals.
The inspector reviewed the applicable licensee documents and it
appears they have taken adequate actions.
-
- -
20
This item ir. closed.
b. (Closed) I.E. Information Notice No. 86-56: Reliability of Main
Steam Safety Valves (MSSVs)
This notice provided additional notificacion of NRC's concern for
the reliability of spring-actuated main steam safety valves
following reports of multiple failures during testing and problems
during power operations and scram recovery.
In memorandum SRT 86-148 of October 3, 1986, the licensee
acknowledged that the problems identified for MSSVs were' applicable
to their valves, and they were working on the issue. To improve
blowdown performance of these valves _ Dresser had issued
recommendations which established new ring settings for 3707 and
3777 valves, which the licensee was incorporating into Procedures
NT.004 and H.25. The licensee considers that their continued
involvement in the B&W owner group secondary relief project will aid
in their effort to improve pressure response on the secondary side. *
The valve performance after resetting these valves will be evaluated ,
by the licensee later during power operations.
Based on the inspector review of the above information and other
applicable documents, it appears the licensee has taken appropriate
actions.
This item is closed.
c. (Closed) I.E. Information Notice No. 86-63: Loss of Safety
Injection (SI) Capability
This notice alerted recipients to a potentially significant problem
pertaining to the loss of SI capability as a result of common-mode
failure of SI pumps from crystallization of boric acid or gas
binding of the pumps. Leaky valves in the discharge line of the
boron injection tank (BIT) could enable highly' concentrated boric
acid to flow through the low pressure discharge line (SI pump
suction) and to precipitate in the pumps which are not normally heat
traced.
In memorandum No.66-304 of October 29, 1986, the licensee stated
that crystallization of boric acid in SI pumps (HPI pumps) may pose
a concern for Westinghouse design plants utilizing highly
concentrated boric acid solution (20,000 ppm), which crystallize at
126*F, but this is not a problem at Rancho Seco. At Rancho Seco,
being a B&W design, the highest boron concentration expected in the
system at BOL is approximately 1400 ppm. .At Rancho Seco the highest
concentration of boric acid solution utilized, is in the
concentrated boric acid tank (CBAT) and it would normally be 8500
ppm (which solidifies at approximately 40 F).
The possibility of concentrated boric acid solution leaking past
normally closed valves and precipitating in the HPI pump (which are
not heat traced), in sufficient concentration to result in
.
.
. . .
21 '
crystallization at ambient temperatures, is ~not considered credible
by the license.
The inspector reviewed Technical Specifications A.12 (Reactor
Coolant Chemical and Hydrogen Addition System), SP 203.02A(B) and
other applicable documents, and it appears that the licensee has
appropriate instructions issued to prevent the identified loss of
safety injection capability.
This item is closed.
d. Review of Licensee's Program to Review Information
Notices /Information Bulletins
The inspector reviewed the following interdepartmental procedures
- (IDPs) and the draft of a new procedure
'
(1) IDP No.-001, Coordinated Commitment Log (CCL)
(2) IDP No.-002, Control of incoming regulatory correspondence
(3) IDP No.-003, Control of outgoing regulatory correspondence
(4) A draft titled - Commitment Management (Issued December 15,
1986)
According to a licensee representative, the new draft procedure
titled: " Commitment Management" will supersede the IDPs noted
above, when issued. These IDPs provide instructions for review,
distribution, and scheduling of performance of corrective actions as
required. In a majority of the Information Notices and Bulletins
reviewed, the actions taken by the licensee appeared reasonable and
appropriate to the substance of the information documents. There
were some I.E. Information Notices, such as No. 85-23: inadequate
i
'
surveillance and post-maintenance and post-modification system
testing (issued March 22, 1985) which the licensee 'was unable to
provide any documentation of what its status was at the time of this
inspection. The licensee could identify who had been ' assigned
responsibility, but that person did not have status within his
group. The inspector stated that licensee management would be
prudent to ensure that their new control system screened new items-
for priority and then periodically tracked action being taken'to.
address concerns.
'
5. Licensee Action on Generic Letters ,
a. (Closed) Generic Letter No. 85-05: Inadvertent Boron Dilution
Events
This letter informed the licensee of the NRC position, resulting
from the evaluation of generic issue 22 (Inadvertent Boron Dilution
Events), regarding the need for upgrading the instrumentation for
detection of boron dilution events in operating reactors. There
_ -. _, _. _ , _.
= y'
. .
22 ,
a
l
were concerns addressed regarding the lack of' distinct, positive /
alarms to alert the operators to boron, dilution events. v,
The licensee performed an analysis of this event and stated, " Based
on the analysis performed in the Updated Safety Analysis Report
~ (USAR), the probability of an unmitigated boron dilution event '
.
'
occurring at this facility is minimal.
The inspector reviewed the USAR Section 14.1.2.4'(Moderator Dilution
Accident); licensee operational analysis (attached to memorandum
EQC-85-427, dated April 29, 1985); Revision 20 to Procedure A.1
which added new substeps 6.1.1.1,_7.1.1.1, 7.2.1.1. and other
applicable documents. The new substeps added this following
information, " Place shif t ' supervisor's clearance on RCDST pumps,
P-622A and B breakers 2E 116 and 2C 516, to prevent boron dilution N
of the RCS when drained down. If these pumps must be run for freeze l,
protection, potential RCS boron dilution should be considered." /
, -r
Based on the information reviewed above, it appears the licensee has f
completed his actions for this item. ,
"
This item is closed. .
o. (Closed) Generic Letter No. 85-20: Verify Stress Analysis
Performed on Modified Thermal Sleeve Designs for HPI Nozzle
In 1982, inspections at B&W plants revealed that some of the high
pressure injection / makeup (HPI/MU) check valve, valve-to-safe-end
weld, safe-end and thermal sleeves'were cracked. A safe-end task
force was formed by the B&W owners' group, which issued a report
with its findings and recommendations to aid in resolution of
generic issue 69. The NRC reviewed the task force recommendations
and agreed that certain actions should be taken. One of the actions
was to perform a detailed stress analysis of a nozzle with a
modified thermal sleeve design to justify long-term operation.
In this letter issued November 8,1985, recent review of operating
experience for some B&W plants has, indicated that the expected -
fatigue analyses could be substantially exceeded by_the end of plant i
life. For example, an increased number of HPI actuation transients g
could occur due -to, manual. actuation. after reactor trips to avoid
~
losing pressurizer level. Therefore, the NRC has determined that it
is necessary that the licensee ensures that valid stress analys6s
have been performed. Each licensee was requested to verify that a
valid stress analysis had been performed for HPI/MU nozzles.and that
the cumulative fatigue usage for these nozzles is ,within the
allowables based on a realistic projection of the_ thermal cycles g
expected for the' life of the plant.
The licensee is tracking the requirement for verification of a valid
stress analysis under item No. 20.0219A, in the Quarterly Tracking
System (QTS), with a due date of December 31, 1987.
> ,
,
L_ _ . _ _ . . _ _ _ _ _ _ _ _ . . _ _ _ . _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . . . _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ - . _ . _ _ _ _ _ . . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _
-
__ _ _ _ - - - - - . _ _ _ _ - - -
,
'
-
23
The inspector _ reviewed the applicable licensee documents on this
subject and it appears that the licensee is taking responsible
actions in following up on obtaining verification of a valid stress
analysis. Based on the licensee commitment to verify the stress
analysis, this item will be closed.
This item is closed.
6. Licensee Action on Part 21 Items ;
a. (Closed) Part 21 No. 85-24: Oil Level Device on Auxiliary
Feedwater Pump is Not Reliable
The licensee issued letter RJR 85-545 on November 8, 1985, stating
that the porthole oil level gauge for auxiliary feedwater pumps
manufactured by Babcock and Wilcox Canada, Ltd, were not reliable
indicators. The licensee stated the gauges were designed with a
metal insert behind the sight glass that could cause the oil to be
trapped so that a false level of oil was indicated. As an interim
corrective action, the licensee recommended removal of the metal
insert. As a long-term corrective action, the recommendation was to
install an oil level sight glass.
The licensee issued ECN No. R-0173 to install a new vertical level
oil gauge to isolate the effect of surface perturbation on the level
indication. This ECN was voided later because the Plant Review
Committee (PRC) did not want any gauge connections on the outside of
the pump bearing housing, that could cause a loss of oil, if
damaged. The PRC considered that the initial interim corrective
action of removing the metal insert, provided an acceptable oil
level reading.
The supplier of this type of oil level gauge (bullseye) indicat.?
that the insert was used as a reflector when observing "Hard-to-see
fluids," and removal of the insert would not cause difficulty in
reading the oil level. Since the metal insert was not shown in any
, pump drawing or technical manual, the licensee considered it an
! optional item and removed it from the applicable sight gauges.
4
Based on the inspector's review of the above information and
applicable documents, it appears the licensee has taken adequate
corrective actions for this item.
This item is closed.
b. (Open) Part 21 No. 86-13: Anchor / Darling-Missing Lock Welds
on Internal Components of Swing and Tilting Disc Check Valves
The licensee received the three letters from Anchor / Darling
identified below:
.. . . . ~ - _ - .. . _
<
'
'-
,
, s
, p. ~
24
4
^
2 '(1) . Anchor / Darling to N. Bradford, Contract Administrator,' dated . ,
-
July 30, 1985, on the subject of cracked tack welds .which lock.
the hinge pin busing in place on the tilting dise-check valves.-
(2) Anchor / Darling to N.' Bradford, Contract Administrator,' dated
!- '
July 31,.1985, on the subject of missing lock welds on hinge ,
pin set screws .of swing check valves at Palo Verde. Nuclear .
~
.. Generating Station.
(3) Anchor / Darling to N. Bradford',TContract Administrator,' dated-
. December 11, 1985, on'the subject of lock welds also missing at ,
i
the hinges support / hinge. support capscrews; interface and at the
hinge support / bonnet interface on swing check valves.
~
-
-
4
+
After reviewing the th'ree Anchor / Darling letters, the licensee <
performed operational assessment 86-8-(signed out June 21, 1986) and -
, generated attachment 1 of that document which identified those <
'
valves that-could be potentially defective..' Assessment 86-8
reconnendations were to implement an inspection / repair program for
the identified valves prior to restart.
The inspector reviewed Operational Assessment,86-8-and requested an
inspection status on the valves in the inspection / repair program.
i. After comparing the valves identified in attachment <1:and the
inspection status report to the list provided with the: .
Anchor / Darling letter of December 11, 1985, the inspector identified. .
to the licensee that it appeared they had failed to include Valve-
SFC-002 (A/DV Assembly.DWG 1338-3) in the inspection / repair' program.
The. licensee representative agreed with the inspector and stated '
that WR No. 120092 would be issued to include ~this valve inuthe
inspection / repair program. ' '
.
t t >
,
As of the date of this inspection,.six palves of the twenty-two' '
valves (m + includes valve SFC-002) identified _in,'the ' - ~
.
inspection /rci program had been' inspected. There~ are seventeen ' -
,
-
swing check valves and five tilting. disc check valves in this; . .
'
program. One of the four swing check. valves inspected required two
. new welds between the. hinge support / bonnet interface and both'of the
tilting disc check valves inspected had hinge pin bushing' retaining
.
,
i welds cracked, and new bushings ~of a new design were installed with-
weld:. .
Based on the inspector's review of the above information, the.
l addition of valve SFC-002 and other applicable licensee. documents,-
l it appeared the licensee has an informal inspection / repair program -e
'in operation that should resolve the Anchor / Darling swing and
tilting check valve concerns about missing lock welds on internal ~
components, prior to the restart of.the unit.
l- .
This item and item 6.c below are unresolved pending review of the-
adequacy of the licensee's reporting of the defects;1dentified. ,
> -
.
-
,
,V-
< ,
.
,
,,
4 ;d * . -#
,. _. -
.
- '-
-v 25
c. Inspector Observation of Licensee Program for Part 21 Items, Not
Generated by the Licensee
It appears the licensee is having some trouble in this area
identifying where these items are received in their organization and
then getting them into a formal tracking system for review and
action as required. When the inspector requested status on Part 21
items No. 86-15-P, 86-22-P and 86-25-P, a licensee representative
stated one had been assigned with no due date and the other two were
not assigned yet. While these Part 21 items are relatively new,
they concern Limitorque valve operators and the. licensee-is now
performing major inspections, repairs and maintenance on Limitorque
valve operators. It may be that the actual licensee personnel
performing the Limitorque work are knowledgeable of these Part 21
items, but it is not identified formally by the licensee.
It appears this area requires additional management attention to
ensure that a Part 21 item is not received by_the licensee, and then
lost or delayed in the system while inspection or work is being
performed on the identified equipment. Once a Part 21 is received, '
it would appear prudent to immediately review it' to see if it'
requires immediate action, or to determine if.it effects existing
inspections or work being performed. Discussions on this subject
were held with licensee representatives, and they agreed this was a
problem and stated they were already working on this. However, no
clear action plan with dates for corrective action had been >
developed. The licensee's resolution of this issue will be reviewed
when the unresolved Part 21 Item 86-13 above is resolved. -
.
7. Exit Meetings
Exit Meetings were conducted on November 21, 1986, and December 12, 1986,
with licensee representatives identified in paragraph 1. The inspectors
summarized the scope of these inspections-and findings as. described in
this report.
!
- __
,