ML20127A414
| ML20127A414 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 06/06/1985 |
| From: | Johnson P, Narbut P, Willett D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20127A403 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 50-206-85-13, 50-361-85-12, 50-362-85-11, IEB-79-27, IEB-80-06, IEB-80-6, IEB-82-02, IEB-82-2, IEB-83-06, IEB-83-6, IEB-84-02, IEB-84-2, NUDOCS 8506210134 | |
| Download: ML20127A414 (13) | |
See also: IR 05000206/1985013
Text
,-
.
,
- -
o
4
,
'
.
'
.
u
U. S. NUCLEAR REGULATORY COMMISSION
.
REGION V
Report Nos.
50-206/85-13,50-361/85-12,-50-362/85-11
Docket Nos.
50-206, 50-361, 50-362
License Nos.
Licensee:
Southern California Edison Company
P. O. Box 800
2244 Walnut Grove Avenue
Rosemead, California 91770
Facility Name: San Onofre Units 1, 2 and 3
Inspection at: San Clemente, California
Inspection conducted: March 4-8, 1985 and April 1-5, 15-19, 1985
)
%lJt
5/31/8 5
Inspector:
P. P. Narbut, Project Inspector
Date Signed
M)&
5 $/~~ S$
.
J. Willett,
eactor Inspector
Date Signed
Approved By:
M
[
P. H.
nson, Chief
Date Signed
Reacto Projects Section No. 3
Summary:
Inspection during period of March 4-8 and April 1-5, 15-19, 1985 (Report Nos.
50-206/85-13, 50-361/85-12 and 50-362/85-11)
Areas Inspected:
Unannounced inspection by regional inspectors of unresolved
and followup items, and licensee actions on IE Bulletins and TMI Action Items.
The inspection involved 86 inspection hours by two inspectors.
Results: One violation was identified regarding an unauthorized design change
(paragraph 4).
.
8506210134 850607
ADOCK 05000206
G
!
m
_'; -
- ,
.-
,
'
-
,
'
., _
-
.
1
.
,
)$
_
4
[-
-
- .; .
,
L
DETAILS
-
,
-
,
c5
.1.
-Individuals' Contacted
a.
- Southern' California Edison Company (SCE)
,
'
.:
.
.
",
. Kr.*J.:G._Haynes, Station' Manager
>
,
J
J*R. W. Kreiger, Deputy Station Manager
.
,-
-
' *D. B. Schone, Site QA Manager
' *P. ' A'. Croy, Manager, Compliance
-
.(*D. E. Brown,. Supervisor, Maintenance Codes.and Welding
,
- R. F. Penn, St'ation Codes Engineer
. .
- D. Sheridan,-Supervisor Plant Maintenance, Unit 1
- H.'. Newton, Manager Maintenance Engineering
'
'
- T.7 . Mackey,. Supervisor Compliance
A.~
-
- C..A. Kergis,: Compliance Engineer
'
F. Briggs,'NSSS,' Mechanical Supervisor
,
'*J. Reilly, Technical Manager
'
..
_~
s
't'
.M. Wharton, Supervising Engineer, NSSS Support
,
v'J. Boardman, ISI Engineer
'
.
'
'
W.~ Savage,. General Foreman, Maintenance Planning
/
- W. M. Lazear, QA Supervisor-
, ,. .
t
~ ~
~
~'
4
,
jL. Rice,, Supervisor,, Site Supply 7
,
P
- w Denotes those. personnel in' attendance at the exit.. interview on April:
'
19, 1985.
-
2.
Licensee Actions on Bulletins and-Previously Identified Inspection Items
~
.
~
.(Closed) IE Bulletin 79-27,: Loss of Non-Class IE Instrumentation
,
- a .-
-
- -and Control Power System Bus During Power Operation', Unit 1
s
a
,
.
[
'
' '
-Background
<
.
,
~
~
,
.
.
,
- Thebulletinrequiredlic[enseesttoreviewpowerlsuppliestosafety
'
and non' safety related instrumentation'and control' systems which
could affect the ability,to achieve cold shutdown,3 to" describe any
resulting design modifications.and*to p'r'epire emergencyLprocedures
~
~
J
.
for such an event. The' licensee'ssactions.for thisibulletin were
~
examined in Inspection' Reports'50-206/80-16, 80,-19,~80-28 and 80-31.
'
Theremaining.actionw'as'toverify'thatthe/ licensee lhadprepared
,'
- emergency procedures for'the-loss of the'125v.DC busses prior to the
~
_
JUnit 1. return to power in;1981. Unit!I returned.to" power on June-
"
"
~
'
17, 1981.
' ' ,
L
b
-This Inspection
"
-
'The.' inspector examined the current' emergency procedure'for the. loss
-
^of the-125v DC buses. The current procedure is' Abnormal Operating
Instruction S01-2.6-4,, Revision 0, dated NovemberL17, 1983. The
l procedure contains information, as required by the bulletin, to
,
,
J
4
--,
f;
~
- 2
,
-
.
W
-
enable the operator to recognize the symptoms of a -loss of the buses
~
and describes the required operator: actions. The inspector verified
that the procedure had in fact .been . issued prior to the 1981 return
to power. The procedure atLthat time was S01-1.7-3,. issued May 4,
1981, and contained ,the same ,information which was reorganized and
later. issued as th'e superseding (and. current) procedure S01-2.6-4.
1
l
.
.
<
Based on the licensee's actions this bulletin is considered closed.
,
b.
(0 pen) IE Bulletin 80-06, Engineered Safety Features (ESF) Reset
Controls
,
The licensee actions regarding the subject bulletin were examined in
report 50-206/81-29. All aspects were closed except valves MOV-
1100B, C.and D which did not respond properly after the ESF signal
was reset. The-licensee had established administrative (procedural)
controls for an interim control measure. A letter from SCE to NRC
dated February 27, 1984 defined the commitment to rework the control
circuitry for the valves during the Cycle 9 refueling in 1986.
This item remains open pending licensee action on valves MOV
1100-B,C, and D.
c.
(Closed)-IE Bulletin 82-02, Degradation of Threaded Fasteners
Background
The bulletin addressed corrosion and cracking of reactor coolant
pressure boundary fasteners and imposed inspection requirements
beyond those already addressed in inservice inspection (ISI)
requirements. The licensee's actions and responses to the bulletin
were addressed in inspection report 50-206/84-11 which closed
certain items. The remaining licensee action was to inspect
specified fasteners when otherwise removed for maintenance or repair.
and report the results to the NRC.
The licensee had requested certain exemptions from the Bulletin
requirements as follows:
Results of fastener inspections will be included in ISI
inspection reports available for NRC review at the site, rather
than reported separately to the NRC as required by the bulletin
(reference SCE to NRC letter dated April 3, 1985).
Nondestruction examinations will only be done if a fastener is
to be reused and if-it has significant degradation, otherwise
only a visual inspection will be performed, unless an
evaluation of the visual examination otherwise dictates
(reference SCE letter dated April 3,1985).
_ _ _ _
_ - _ _ - _ _ _
_
.
3
.
.
This Inspection.
_
The inspector examined the procedures in place to perform the
bulletin required inspections. The procedures ' applicable are:
- -
'S0123 XVII-1.0, Revision 0, dated October 3, 1984, " Inservice
Inspection Program Implementation"'
.
S0123-I-7.19, Revision 9, date'd August 24,.t1983, " Monitoring of
Threaded Fasteners (Bolts / Studs) in the Reactor Coolant
Pressure Boundary (RCPB)"
,
Theprocedureswerefou$dtoadequatelydefinethebulletinrequired
actions for fastener inspection.
The inspector examined the licensee's inspection data for bolting
from the following components.
Pressurizer Relief Valve bolts for valves 533 RV, 532 RV
Steam Generator manway bol'ts for steam generators A, B and C,
hot and cold leg sides
RHR valves MOV 833, MOV 813, M0V 814
'
The bolt inspection results were satisfactory for the pressurizer
code safety valves and the steam generator manways. The inspector
verified that data were recorded and actions were taken in
accordance with procedure requirements.
The results for the RHR valves were unsatisfactory. The licensee
had performed a walkdown for corrosion as part of the return to
service effort for Unit 1.
That walkdown identified corroded
fasteners on the RHR valves. The bulletin's fastener inspection
requirements were properly invoked by the licensee maintenance
personnel in conjunction with their decision to replace the
fasteners.
As will be discussed separately in paragraph 4 of this report, the
selection of replacement fastener material was not proper; a lower
strength material was used without proper engineering review and
approval.
In regards to bulletin actions however, the fastener inspection data
for the corroded RHR valves also appeared to be improperly taken.
Specifically, the maintenance orders required inspection of the
fasteners (in accordance with the procedure) by visual examination
and then a surface examination by NDE. This was required "even
though existing fasteners may be replaced with new fasteners." The
purpose of the bulletin was to examine corroded fasteners for
evidence of stress corrosion cracking by performing surface
examinations.
E
-.
-
-
-.
-.
,
.
.
. . _ . ~
7
-
.-
- y
.4
~
'
'
,
.
,
,g
lTheLprocedural step signoffs by the maintenance' mechanics (on the-
maintenance' work orders))were marked."N/A" in the steps for
-
.
- y
decontamination of the bolts,. examination of'the bolts, and
h
- forwarding-of the results'to' engineering for evaluation. The steps-
~
-
- were annotated that the
- step was N/A'd because the bolts could not
'
' ' .
'be decontaminatedLand were too corroded for inspection. -The.
'
.,
.
-inspector considered .that, technically, the rationale for not
_
e performing the . inspection was poor. The licensee has facilities for
,
performing' inspections of contaminated-items. Additionally,-the
- corroded condition of the bolts made.them good examples of the kinds
L,
'
.of~ conditions the bulletin inspections were directed towards,..rather:
,
~
than a reason not_to perform an inspe'ction.
s
,
.
The maintenance. orders involved.were:
U
'MO 306819 for MOV 813
.
'.
1*
MO 307501 for MOV. 814 '
'
"
MO 84042175001 andsMO 307503 for'MOV.833c4
'
1
Thecompletedmaintenanceobder.(with.tbeinappropriatelyN/A'd
steps) was reviewed and' approved by the procedurally required
organizations which should have flagged ~the; inappropriate decision.
's
.These organizations inc1'uded QA-and Engine'ering'. The inspector
"+
considered the issuance of a' Notice of Violation -for'this matter but
,. concluded a violation?was not involv,ed for the following= reasons:
'
,
- -
. Lack of Safety Significance: The/ fail'ure to inspect bolts did-
1
h
not meet procedure requirements'atf,the, time, but'do meet
'.
current requirements ~('einspect for. reuse ~only). Secondly,
r
'
. there has not been a history of stress corrosion cracking of
>
this material. The bolts were. examined by.the licensee'(as a
. catch-up-effort during this' inspection) and none had stress
,
corrosion ~ cracking problems.
.
'
Previous Licensee Actions: As'a result'of other' violations the
'
licensee had since taken. aggressive actions regarding the."N/A"
of procedural steps. These actions were'not.in place at the-
time of the Solt inspections in 1984.
Conclusion:
Based on the licensee's sctions and the revised requirements
'
reflected-in the licensee's April 3, 1985 letter.to the NRC which
reduce the data gathering and reporting requirements, this. bulletin
is considered closed.
.+
d.
l(Closed) IE Bulletin 83-06, Nonconforming Materials Supplied by
'
g }lC
Tube-Line Corporation
- ,
"
.Backgroundi
The-licensee's actions for this bulletin were previously examined in
inspection. reports 50-206/84-23 and 84-30. At that time the
. .
inspector'had found additional installed Tubeline material which'had
'
>
r
..
A
u
m
. -
. .
.
.
. . .
-. . -
.-
5
-
.
not been identified by the licensee's search. The remaining actions
were for the licensee to submit a revised response to the bulletin,
to include ~ actions to prevent recurrence of an incomplete search and
to ensure previous actions taken regarding generic communications
~
,
(on material problems) were properly investigated'and evaluated.
This Inspection
^
'
The inspector reviewed the licensee's revised response to the
bulletin. The revised response was dated December 4,1984. The
licensee identified the reason for the initial incomplete search as
having failed to review the material provided through Bechtel field
purchase orders. This search was a general one (for material
searches) during the period of September 1980 through September
1984, when no Bechtel field organization was in place at SONGS 1 to
review field purchases. During this time the Bechtel search mode
was limited to home office purchase reviews.
It should also be
noted that the specified scope of the bulletin was ASME piping, and
the Tubeline fittings found at San Onofre 1 were in systems not
constructed to the ASME Code.
To prevent the recurrence of incomplete searches, the licensee has
initiated significant action in the form of a program called COPE
(Control of Problem Equipment) which is planned to be totally in
place by June 1985. The program was generated in response to the
inspector's findings regarding this bulletin and Bulletin 84-02
(paragraph 2.e) regarding HFA relays. The completion of the COPE
program will be examined as part of the followup of that Bulletin.
Additionally, SCE instructed Bechtel to perform a reinvestigation of
material searches conducted during the period that a Bechtel field
procurement group was not in place.
SCE audited those Bechtel
remedial actions and found them satisfactory. Additionally, Bechtel
identified, in a letter dated January 11, 1985 (McClusky to Nunn),
how they would revise and proceduralize future material searches to
ensure completeness.
Subject to the followup of the licensee's new COPE program to be
followed up as part of Bulletin 84-02, this bulletin (83-06) is
considered closed.
e.
(0 pen) IE Bulletin 84-02, Failure of GE HFA Relays
Background:
The licensee actions for this bulletin were examined and described
in inspection reports 50-206/84-30 and 84-33. The inspector had
determined that the licensee's material search had not been
comprehensive; the inspector discovered the subject relays in the
warehouse. The remaining items regarding the licensee's actions
were verification of the COPE program (Control of Problem
Equipment), and verification of the procurement Quality Assurance
surveillances of the adequacy of licensee actions for previous
w
-
-
.
-
-
. -
-
-
.
-
-
-
-
-
- -
e
6
-
.
-
<
.
material searches conducted for IE Bulletins and Information
Notices.
This. Inspection
'The inspector reviewed site order S0123-CP-1, Revision 0, dated
February.27, 1985, which outlined the program to be established and
defined responsibilities.
The inspector met-with responsible personnel and reviewed the
schedule, which shows-full implementation by June 1985.
-This bulletin remains,open pending' implementation of the COPE
program and ' review of the procurement QA verification actions.
3.
Licensee Actions on Previously Identified Inspection Items
-a.
(0 pen) Followup Item 50-206/80-11-03, No cold-to-hot
relief pressure correlation factor for the Pressurizer Code Safety
Valves setpoint testing
Background:
The inspector had previously. observed in the 1980 inspection that
the licensee performed required periodic set pressure tests of the
Unit.I' Pressurizer Code Safety valves in a cold condition.
Cold
testing of safety valves is a generally acceptable method of
verifyingfset pressure provided that the expected difference in the
pressure at which the valve actuates (in the hot versus cold
condition)'is known and accounted for.
This requirement, for.a cold-to-hot correlation factor, is derived
from the Technical Specification requirement to perform periodic
relief pressure tests.of the subject valves in accordance with the
requirements of ASME Section XI. The ASME Section XI requirements,
,
in turn,' reference ASME PTC 25.3 for the' performance of safety and
L
relief valve testing. The. introduction of PTC.25.3,~Section.0,;
cautions that""...if the~ temperature'of the medium used to test the
valve differs substantially- from th'e te'mperature to which the valve
i, subjected while in service, the, opening and closing
prassures. . .will be different from the test pressure. . .it ~is
naressary to develop ap'propriate corrections for the' valve under
' test to account for thesefdifferences.",'i.;
.
c
.
3
-
~-
,
The open status of this item was discussed at 'a regional inspector's
exit interview on December: 21,'1984.
This Inspection
The inspector reviewed the current versions of the maintenance and
test. procedures applicable to the pressurizer code safety valves
(PRV 532 and 533) and confirmed the licensee had not yet included a
cold-to-hot correction factor in'their set point test procedure.
-
_ . _ - - _ , ~ - .
-,_
-
'
7
,
..
s
y
The current set point e'st procedure is S01-1-2.3, Revision 2, dated
October 14, 1984. The valve is tested cold with a nitrogen medium.
'
The inspector met with t.he responsible site engineering manager and
his technical' staff to discuss'the item. The staff presented
'information which demonstrated that the issue of: proper testing of
safety and relief valv'es. is an-industry wide subject of interest.
There have been several initiatives by various organizations aimed
'
at resolving the issue. ,Two notable efforts are (1) an NRC
~
requirement (imposed in the technical specifications) for newer
plants to test safety valves hot,and with the same medium as seen in
- service and (2) an ASME code committee'(OM-1) effort to revise the
code requirements to require essentially the same thing but provide
some' options. The licensee had gathered some data, from TVA and
'EPRI, but the data showed significant scatter.
The data do not
apply directly to the Unit 1 Crosby valves but are consistent in
that the hot tested' set point is always lower than the cold tested
setpoint (the co'nservative direction).
The inspector-requested the licensee to provide rationale to
~
' demonstrate that the change in set point (from cold-to-hot) for the
Unit I safety valves would be in the conservative direction and to
affirm that'no safety analyses would be adversely affected by the
expected change in set point. The licensee presented general
industry information which demonstrated that safety valves have a
hot set point.from 1 to 180 psi. lower than the cold set point. The
'
fact that San Onofre does not have a history of inadvertent relief
valve lifts indicates that the change in set point for a hot valve
is not an operational problem at normal operating pressures. The
'
licensee'also presented a memorandum dated A,ril 16, 1985 from the
Westinghouse site representative to Station Technical, SCE. The
~
memorandum states:
Westinghouse has no directly applicable data for the Unit 1
valves but the same type of conservative change can'be
expected.
There is no unresolved safety concern with a cold-to-hot set
point drift (in the lower set point direction).
At the exit interview the inspector discussed the findings regarding
this item, specifically that no apparent action toward resolution
had occurred since the item was identified in 1980 and that the
information provided indicated there was not an unresolved safety
question involved. The inspector requested the licensee to identify
the path to resolution of this item.
Licensee management stated
that the resolution of this item would be addressed in the context
of a larger initiative, that being the licensee's plans for a
general revision and reformatting of their technical specifications
to be accomplished in conjunction with their SEP actions.
Specifically, the licensee plans to update the FSAR (and technical
specifications) after the completion of the SEP evaluation
(scheduled for 1985) in accordance with 10 CFR 50.71(e)(3)(ii).
c
8
'
.
.
,
This item remains open'ending 'further licensee action.
p
4.
Maintenance Issues
During the examination of IE Bulletin 82-02'regarding degradation of
fasteners, the inspector examined certain aspects of the licensee's
controls for maintenance and maintenance testing. These aspects included
~
compliance with ASME Section XI repair and testing requirements, material
control, engineering involvement.in design changes, records' accuracy,
procedure adequacy, torquing requirements, proper test pressure, and
supplier material certifications. Specifically, these aspects were
reviewed as they related to the replacement of corroded carbon steel
fasteners with stainless steel fasteners on RHR valves MOV 833, 813 and
814. All aspects examined were found satisfactory with the following
exceptions:
Improper Design Change
The maintenance orders issued to replace corroded body to bonnet studs on
the RHR valves replaced the original carbon steel fasteners with lower
strength stainless steel fasteners. There were two central problems
identified by the inspector:
No design analysis was performed to assure the lower strength
material was adequate for service.
Licensee personnel failed to recognize the material substitution as
a design change.
The specific details involved are as follows:
The valves involved are RHR MOV-813, 814 and 833. The valves are reactor
coolant pressure boundary components. Valve 813 is the first isolation
valve in the reactor coolant system loop C outlet to the RHR system; and
therefore sees RCS pressure in service.
The associated documentation for the valves is:
Valve
Maintenance Work Order
ASME Section XI Abstract MERS*
MOV 813
306819
S01-011-83
010-83
814
307501
S01-009-83
007-83~
833
307503
S01-010-83
008-83 Rev. 1
Maintenance Engineering Repair Specification
The original construction design code was ANSI B31.1, 1964 Edition, but
repair and replacement are to be done to ASME Section XI,1977 ' Edition.
The manufacturer was Crane Valve Company, drawing DR 33473 (for 6" valve
833) and drawing DR 33463 (for 8" valves 813 and 814).
The original bolting material specified by Crane was ASTM A-193 Grade B7
Chrome-Moly Carbon steel (which has a minimum tensile strength of
approximately 125 ksi). The replacement material installed by SCE was
r-
9
..
4
ASTM A 193 Grade B8 Chromium Nickel Stainless steel (which has.a minimum
tensile strength of approximately 75 ksi).
The responsible station engineering manager stated a design
reconciliation analysis had not been performed to verify that the
material substitution was technically adequate. During the inspection,
preliminary analysis was performed in response to the inspector's
findings and the preliminary results showed that the new material was
satisfactory for service, that design margins have possibly been reduced,
but that there was no cause for an immediate safety concern. The
responsible engineering manager so stated ~during the exit interview on
April 19, 1985.
The Code of Federal Regulations, 10 CFR 50, Appendix B, Criterion III,
state in part that:
" Design changes including field changes shall be subject to design
control measures commensurate with those applied to the original design."
The licensee's design control measures for repair and replacement
activities are specified in the SCE Topical Quality Assurance Manual,
Appendix IV "ASME Code Section XI Repair and Replacement Program" which
states in part " Repairs and replacements performed at Unit I will
implement the requirements of ASME Code Section XI, 1977 Edition through
Summer 1978 Addenda...."
~
ASME Section XI, IWB 7600 states in part " Materials shal1 comply with the
requirements to which the original component or part was constructed."
IWA 7210 allows for replacement material provided:
" Modified or altered
designs are reconciled with the Owner's Specification through the Stress
Analysis report, Design Report or other suitable method whi-h
demonstrates satisfactory use for the specified design and operating
conditions...."
Contrary to the above, carbon steel body to bonnet studs were replaced
with significantly weaker stainless steel studs on RHR valves MOV 813,
814, and 833 without a design reconciliation by a suitable method which
demonstrated satisfactory use for the specified conditions.
This is an apparent violation.
(Violation 50-206/85-13-01)
The inspector also-noted a similar carbon to stainless bolt material
substitution was authorized in 1981 for pipe flange bolting. Flange
bolting generally sees a lesser loading than body to bonnet valve bolting
and is generally very conservative in design strength. There was no
evidence at the time of inspection that the material substitution
authorized was, in' fact, implemented. Therefore no violation was
identified. The material substitution was authorized by a change to the
" Piping Design and Material Specification" Number M-18668 Sh 179 Revision
0, Configuration Change Notice 1 dated January 19, 1981.
At the exit interview the licensee representatives committed to determine
if the material substitution authorization was implemented and to verify
t
c.
-
.
~
10
that the substitution authorization is a technically sound design change.
'(Followup item 50-206/85-13-02)
Improper Test Pressure
RHR Valve MOV 813 is a boundary gate valve between'the RCS and RHR
systems. As such, in normal operation it is shut and sees RCS normal
operating pressure rather than the much lower normal operating pressure
of the RHR system on its "high pressure" side and in the valve bonnet
area. Upon' replacement of-studs the ASME Code requires a leakage test at
"not less than the normal operating' pressure associated with 100% rated
reactor power".
The studs in questions see RCS system pressure in service but were leak
tested at the lower RHR system pressure.
This matter is primarily one of test data record validity since the studs
did eventually see RCS system pressure and normal planned QC walkdowns
would have identified any leakage problems. At the exit interview, the
licensee committed to examine methods to improve the detailed
specification of test requirements for system boundary valves.
(Followup
Item 50-206/85-13-03.)
5.
TMI (NUREG 0737) Activities (Unit 2)
The inspector reviewed the licensee's program for/and implementation of
NUREG 0737 items I.C.1, I.D.2, and II.F.2: Accident and emergency
procedure implementation, safety parameter display system, and
instrumentation for detection of inadequate core cooling, respectively.
a.
I.D.2 (0 pen) Safety Parameters Display System (SPDS)
The accident monitoring system includes the Critical Factors
Monitoring System and the Safety Parameters Display System. A
sub-set to this is the Q-SPDS (SPDS for seismically qualified'
parameters). The licensee's implementation of the requirements in
I.D.2 was as follows:
The operability requirements of the SPDS (required by
NUREG-0696) are included within the technical specification
requirements for the Accident Monitoring System
instrumentation. The SPDS is installed and complete except for
that which is supplied by the heated junction thermocouples
(HJT). The licensee stated that the design change package
which incorporates the HJT's will be complete when the head is
installed, the instruments are connected to the SPDS and the
cables are rung-out, tested, and calibrated. This work is
scheduled to be completed just prior to restart after the first
refueling outage.
The inspector observed the use of the SPDS system by license
candidates in the control room, and verified their training and
understanding of the system.
Closure of this item will be
verified at a later inspection when the HJT work is complete.
p
m
-
A
11
+
,
.. b .
II.F.2 (open) Instrumentation for Detection of Inadequate Core Cooling
Generic Letter No. 83-37 identifies those TMI items for which
technical specifications are required. ~ Enclosure 1, item 10,
provides guidance for II.F.2.
Enclosure 3 provides model technical
specifications. The Subcooling Margin Monitor (SMM), the Heated
Junction Thermocouple System-and the Core Exit Thermocouples (CET's)
comprise the instrumentation for detection of inadequate core
cooling. The licensee's implementation of the requirements of
II.F.2 was as follows:
San Onofre-2 technical specifications (amendment 31) address
some item II.F.2 instruments in numbers 11, 18 of the accident
monitoring instrumentation tables for limiting conditions of
operation and surveillance requirements. On January 11, 1985,
the C.E. Owners Group submitted (after discussions of
fundamental disagreement) to NRR proposed HJT System Technical
Specifications. The licensee has submitted a Technical
Specification proposal (consistent with the C.E. Owners Group
proposal) to NRR for consideration.
Procedure implementation and training for prompt recognition of
inadequate core cooling using existing instrumentation (per
NUREG-0578) was observed during license examinations on the
plant simulator. Procedures used by operators to recognize
inadequate core cooling, which rely on data from the SMM, HJT,
and CET's .(per NUPIG-0694) will be evaluated when TMI item
I.C.1 is reviewed.
This item will be closed out after:
(1) Completion of the;HJT system work
4
-
(2) When NRR has reviewed and amended the~ technic 1 specifications
(3) When a review of selected ' emergency procedures which use SMM,
HJT, CETs, and the SPD System is completed'during a follow-up
inspection of TMI. item I.C.1.
TMI Item I.C.1 (open) Accident and' Emerge'ncy Procedure
c.
Implementation
,
.
This TMI item requires' licensees to:
[
,
y
.
Perform analyses of transients and accidents (including small
break LOCA's and inadequate core / cooling), and prepare
emergency procedures (symptom based) for multiple and
consequential failures.
,
Revise procedures to address inadequate core cooling,
transients and accidents (per NUREG-0694 and NUREG-0578),
maintaining consistency with the Final Long-Term guidance
contained in NUREG-899 (implementing document for-the emergency
procedures pilot program--TMI item I.C.8).
. -
-
-
-
.-
a-.
-
-
-
,
. 7
,
l
. . +
'-
'- 12
. . .
i
Y
s
L
,
..
i
The' inspector verified thelexistence, use and training of personnel,
in symptom based single / multiple ~and consequential transient and
accidentprocedures,during}estingoflicensecandidatesonthe
plant simulator.
.
'
.
.
-
~ Procedures which address inadequate, core cooling will be reviewed
'during a follow-up inspection ~when the HJT System and QSPD. System
,
are; complete-and the technical specifications which address them are
approved by NRR.
<
,
,
-6.
Exit Interview
.
The inspector met with the licensee personnel denoted in paragraph 1 on
April 19, 1985. The inspection details and findings as noted in this
report were discussed.
.
4
7
~p'
l
l
.
l..