ML20050B035
| ML20050B035 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 03/03/1982 |
| From: | Mccabe E, Mccann J, Pullani S, Rhoads G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20050B029 | List: |
| References | |
| 50-387-82-04, 50-387-82-4, NUDOCS 8204020521 | |
| Download: ML20050B035 (19) | |
See also: IR 05000387/1982004
Text
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _
.
U.S. NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
Region I
Report No. 50-387/82-04
Docket No. 50-387
,
Category
B
License No. C_PPR-101
Priority
--
Licensee:
Pennsylvania Power and Licht Cornany
2 North Ninth Street
Allentown. Pennsylvania 18101
Facility Name: Susquehanna Steam Electric Station
Inspection at: Salem Township, Pennsylvania
Inspection conducted: January 27 - March 1,1982
Inspectors:
b
3l
ez
Gary W. Rhoads
date signed
Ns
7 "L
==
John FV M
date signed
'
3l3l82.
.) P
- '
S. V.-Pullani
date signed
Approved by:
DAM
3l3181
Ebe C. McCabe, Chief, Reactor Projects
date signed
Section 2B, DPRP
Inspection Summary:
Inspection on January 27 - March 1,1982 (Report No. 50-387/82-04)
Routine resident (208 hr.) and regional inspection (48 hr.) of: Startup Test Program,
Startup Procedures Review; Preoperational Testing; Preoperational Implementation;
Bulletins and Circulars, Open Items; and Plant Status. Thirteen open items, four
bulletins, two circulars, and one Construction Deficiency Report were closed.
Five
violations were identified; two for improper incorporatica of test requirements in
preoperational tests, one for inadequate control of environmental condition for
activities affecting quality, and two for failure to follow approved procedures.
Six other items were opened during this inspection.
Region I Form 12
'
1
(Rev. April 77)
0204020521 820315
PDR ADOCK 05000387
G
._
_.
.
_
. .
..
.
DETAILS
1.
Persons Contacted
Pennsylvania Power and Light Company
L. Adams, Plant Supervisor of Operations
T. Clymer, Site QAE
F. Eisenhuth, Senior Coapliance Engineer
E. Gorski, Plant Quality Supervisor
J. Green, Operations Quality Assurance Supervisor
H. Keiser, Superintendent of Plant
C. Myers, Assistant Superintendent of Plant - Outages
R. Sheranko, Startup and Test Group Super isor
D. Thompson, Assistant Superintendent of Plant
Bechtel Corporation
N. Covington, Assistant ISG Supervisor
E. Figard, ISG Supervisor
The inspectors also interviewed other PP&L employees, as well as employees
of Bechtel.
2.
Licensee Action on NRC Findings:
a.
(Closed) Unresolved Item (387/80-14-01) Station Quality Certification
Records.
FSAR Table 17.2-1 has been revised to commit to Regulatory Guide 1.58,
Revision 1 which endorses ANSI Standard N45.2.6-1973.
Quality Control Pro-
cedure (QCP) 10, Revision 0 " Training, Qualification, and Certification of
Inspection and Test Personnel" was reviewed for incorporation of the re-
vised Regulatory Guide. The qualification requirements specified in the
procedure met the minimum qualifications as defined in the ANSI Standard.
b.
(Closed) Unresolved Item (387/80-20-08) Core Spray System Drawing in FSAR.
FSAR Figure 6.3-4, Revision 25 was reviewed. All discrepancies previously
identified had been corrected by this revision.
c.
(0 pen) Unresolved Item (387/80-10-01) Acceptance Criteria For Battery Loading.
FSAR Tables 8.3-6A, B, C, D, Revision 19 and Table 8.3-7, Revision 23 were
reviewed to determine if the tables had been updated to show the load cycles
for the 125 VDC and 250 VDC batteries during a LOCA condition. The tables
had been updated, but it now appears that the initial load plac2d on both
the Division 1 and II 250 VDC batteries during preoperational test P88.1
does not meet the load spectrum as shown on FSAR Table 8.3-7.
d.
(Closed) Inspector rollowup Item (387/80-28-02) Comments on Primary Contain-
ment Instrument Gas Test.
Work authorizations WA-U-11179, WA-U-11452, ar.d WA-U-11957 were reviewed.
,
From the review it was detennined the remaining relief valve had been satis-
factorily bench tested.
1
2.
e.
(Closed) Inspector Followup Item (387/79-08-06) Environmental Oualification
c f Stem Mounted Limited Switches.
The licensee has submitted their environmental qualification report to NRC:
NRR for review. NRC:NRR has stated in the Safety Evaluation Report, Supple-
ment 2 that after review they will review the program ard report on its finding
in a future supplement.
f.
(Closed) Inspector Followup Item (387/81-24-06) Conduct of Operations.
See paragraph 4 on Circular 81-02.
9
JClosed) Unresolved Item (387/80-28-03) Remote Snutdown Panel Design.
Based on the inspector's concern identified in this item,NRC:NRR is per-
forming a review of the remote shutdown panel
and has requested the licensee
to make some changes to the panel.
h.
(Closed) Unresolved Item (387/81-02-10) Resolution of Document Review.
Memoranda documanting resolution of coments made by the Station Nuclear
Engineer were reviewed.
During NRC Inspection 81-29 the inspectors had
reviewed recent licensing submissions to the NRC, and had verified existing
procedures for document review is being followed.
i.
(0 pen) Inspector Followup Item (387/80-32-06) Control Rod Drive Hydraulics
Preoperational Test.
On February 4,1982 Preoperational Test P55.1, Revision 2 was reviewed to
ensure proper resolution to NRC inspector comments were incorporated into
the test.
Comments concerning discrepancies between as-built descriptions
of the scram discharge instrument volumes and the FSAR Revision 28 descrip-
tion have not resulted in a change to the FSAR.
Comments concerning drive
water pressure valves did result in a change to the FSAR Section 4.6.1.1.2.4.1,
but were not included in changes to FSAR Section 4.6.1.1.2.5.1.
These
comments will be reviewed in a subsequent inspection.
On February 4,1982, the official Test Copy of P55 .. Revision 2 was re-
viewed.
It was noted that Acceptance Criteria 2(11) stated that total
cooling water flow for control rod drives (CRD) shall be between 37 g.p.m.
and 63 gallons per minute (9.p.m.) and indicated this was verified in para-
graph 7.3.7 of the preoperational test. The step had been signed off as
being verified by the System Test Engineer on January 26, 1982.
Section 7.3.7
was reviewed.
The section did not verify total CRD flow as indicated in the
acceptance criteria, bu+ established flow at 63 ! 2 g.p.m. with no exact
indication of flow being documented for final established flow.
10 CFR 50, Appendix B, Criterion XI states that a test program shall be
established in accordance with written test procedures which incorporate
the acceptance limits contained in applicable design documents.
PP&L Quality Assurance Manual Procedure SP-3, Re'tision 2 states that the
preoperational test program shall consist of approved procedures which assure
data from the test is evaluated against the design requirements to determine
acceptability of test results.
3.
Startup Administrative Manual Precedure AD7.5, Revision 10, states that
acceptance criteria will specify either minimum or maximum values, or
tolerances and that the acceptance criteria will identify steps of the test
procedure that verifies the stated ' criteria.
On February 5,1982 the inspector informed the Superintendent of Plant that
listing minimum and maximum valu~es for total cooling water flow in the
acceptance criteria, but verifying total cooling water flow with tolerances
which put total cooling water fic i above the maximum value stated in the
acceptance criteria was a violat in.
(387/82-04-01)
j.
(Closed) Inspector Followup Item (387/80-24-04) Main Steam - Nuclear Steam
Sunply Shutdown System Preoperational Test Comments.
On February 17, 1982 Technical Procedure TP2.17, Revision 0, and Preopera-
tional Test P83.1A, Revision I were reviewed. All previous inspector comments
have been resolved by these two documents.
k.
(0 pen) Inspector Followup Item (387/80-01-04) Emergency Service Water Pipe
Interferences.
As of February 18, 1982 the Bechtel Quality Assurance Request (QAR-F584)
remained open requesting resolution of the inspector :omments.
1.
(Closed) Inspector Followup Item (387/80-32-05) Steam Relief Valve Acoustic
Monitoring Testing.
On February 18, 1982 Test Change Notice (TCN) numoer 016 to Preoperational
Test P83.1B, Revision 2 was reviewed.
This TCN incorporated testing for the
acoustic monitoring system for the steam relief valves.
(0 pen) Inspector Followup Item (387/80-32-07) Reactor Recirculation System
m.
Preoperational Test Comments.
On February 16, 1982 the inspector reviewed Preoperational Test P64.1,
Revision 3 and the FSAR Sections 7.7.1.3.3.4.7 and 7.6.la to determine if
all previous inspector comments had been resolved.
Two of the previous
comments dealing with FSAR changes had not been incorporated into revisions
to the FSAR.
These comments dealt with a description of the speed limiters
and the RPS breaker trip logic.
On February 17, 1982 two draft FSAR change
request forms were presented to the inspector which appeared to satisfy the
concerns.
This will be reviewed after L aroval of the draft change,
n.
(Closed) Inspector Followup Item (387/81-07-03) Control Rod Drive High Point
Vents.
On February 17, 1982 a PP&L internal letter PLI-13712 was reviewed.
This
letter stated the licensee's NPE group had evaluated the leaking vents and
j
determined the problem encountered to be not reportable.
. - - -
.
.
.
4.
'
o.
(0 pen) Violation (387/80-16-03) Failure To Incoroorate Recirculation
System Flow Testing Into The Preonerational Test Precedure.
On February 18, 1982 Preoperational Test P64.1, Revision 3 was reviewed to
assure that integrated operation of the recirculation system motor generator
sets, recirculation pumps, jet pumps, and valves were incorporated in the
test.
PP&L internal letter, PLI-12299, was also reviewed requesting that
recirculation flows at greater than 45% pump speed be tested during the
startup phase. These changes appear to have fulfilled all immediate correc-
tive action to correct the Recirculation System Preoperational Test.
The licensee stated in its response to the NRC, PLA-559, that to avoid further
items of noncompliance the Test Revtew Board had been directed to insure
that FSAR requirements had been incorporated in the preoperational test.
This corrective step does not appear to have been sufficient since two items
of noncompliance in the present report period (387/82-04-01 ) and (387/82-04-
04) are partially due to FSAR requirements not being incorporated into Pre-
operational test. On February 19, 1982 the Superintendent of Plant was told
th1s item would be closed out based on further corrective action taken to
assure the Test Review Board is satisfying its responsibilities,
p.
(Closed) Inspector Followup Item (387/81-02-02) High Pressure Coolant In-
jection Preoperational Test Comments.
Preoperational lest PS2.1, Draft Revision 2, and FSAR Revision 19 were ob-
served. All inspector corments have been resolved.
q.
(0 pen) Inspector Followup Item (387 80-28-05) Temaorary Modifications In
/
Containment Atmosphere Circulation System Durino
)reoperational Testing.
During NRC Inspection Report (387/80-28) in October, 1980 three Startup
Field Reports (SFR's) Numbers 1187, 1230, and 1531 were noted to be open
for resolution during preoperational testing of Containment Atmosphere
Circulation.
On February 18, 1982 the three SFR's were reviewed to determine if resolu-
tion to the identified problem could affect validity of the preoperaticnal
test. SFR 1230 which documented a problem with the circulation fans tripping
on overload during starting in fast speed resulted in the issuance of
Desion Change Package (DCP) Number 297, Revision 1, which will result in
qualified overloads of higher capacity being installed in the circuitry.
SFR's 1187 and 1531 document.ed discrepancies between wiring terminations
for the containment circulations fans control and alarm circuitry and the
i
design drawing for this circuitry (drawing E-224, Sheet 1).
SFR 1187 was
resolved on August 13, 1980 stating the alarm circuit on E-224, Sheet 1,
would be revised to show circuitry as found in field.
SFR 1531 was issued
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
.
.
,
p
5.
on October 8,1980 to state that although E-224, Sheet 1, had been re-
vised (Revision 10) the circuitry as shown on the drawing still did not
correspond to what was in field. On November 10, 1980, the response from
Bechtel Project Engineering stated that they did not agree that E-224,
Sheet 1, was in error. Attached to the SFR was a memorandum of a tele-
phone call between a member of Bechtel Project Engineering and the Inte-
grated Startup Group Test Engineer for the containment atmosphere circula-
tion system stating that the alarm circuitry shown on drawing E-224, Sheet
1, Revision 10, is correct and no drawing change is needed.
On February 18, 1982 an inspection of the alarm circuitry for the contain-
ment circulation fan IV416A in panel 1B236 was conducted. The alarm cir-
cuitry had still not been changed to correct Ge installation to agree
with the SFR 1531 resolution. The temporary modification log was reviewed
and it was determined that a temporary modification was still in effect
for the configuration found in the panel.
Since the problem identified in SFR 1187 and 1531 still exists, this item
will remain open pending resolution.
On Fetruary 20, 1082 Startup Administrative Manual, A06.3, Revision 6,
"Startup Field Report" was reviewed.
Section 5.3.3 of the procedure states
that upon receipt of the SFR reply, work items resulting from the SFR
resolution agreeable to the ISG are added to the Startup Work List, and
if ISG disagrees with the resolution, a new SFR indicating the disagreement
is prepared and processed. No SFR had been written since the reply of
November 10, 1980 to indicate that ISG disagreed with the resolution, and
no Startup Work List item had been entered as of February 18, 1982.
PP8L's PSAR, Appendix D, Section 0.2.5, the PP&L 0A Manual Procedure 7.1
and 10 CFR 50, Appendix B, Criterion V require that activities affecting
quality will be accomplished in accordance with approved procedures. On
February 25, 1982, the Assistant Superintendent of Plant was told that the
failure to properly follow the Startup Administrative Manual Procedure
AD.6.3, Revision 6 for processing the resolution to an SFR was a violation
against 10 CFR 50, Appendix B, Criterion V.
(387/81-04-02)
,
r.
(Closed) Inspector Followup Item (387/81-28-02) Sources of Acceptance
Criteria.
l
The inspector verified that sources of acceptance criteria are identified
on several of the startup test procedur,es previously reviewed by the in-
spector. Licensee personnel stated that the sources of acceptance
criteria will be identified on all future startup test procedures,
A,
.
_ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ - . _ _ . _ _ _ _ _ _ _ _ _ _ _
.
.
6.
s.
(Closed) Inspector Followup Item (387/81-28-03) Transient Tests on Reactor
Level Instruments.
~~
The licensee personnel stated that the transient tests on reactor level
instrumeats are included in startup test procedures ST-25, Revision 0, Main
Steam Isolation Valves and ST-27, Revision 0, Turbine Trip and Load Rejection.
The inspector reviewed these procedures and verified the above.
t.
(Closed) Construction Deficiency Report (387/80-00-15)
See Section 4 on Bulletin 80-21.
3.
Plant Tours
The inspector conducted periodic tours of accessible areas of the plant during
normal and backshift hours. The inspector observed work in progress, testing,
housekeeping, cleanliness controls, and storage and protection of components
and systems.
Findings:
ESW and RHR Instrumentation Access Pit
a.
On February 1,1982 the inspector found wate r in the two concrete access
pits containing 'B' division Emergency Service Water and Residual Heat Re-
Each pit contains a
moval (RHR) Service Water Piping and Instrumentation.
section of discharge piping with a flow orifice, heat tracing, a flow trans-
mitter and associated wiring and terminal boxes. The 'B' division Emergency
Service Water pit was filled to about 5 feet, submerging all but the tops
of the terminal boxes. The 'B' division RHR service water pit was filled to
about 2b feet, partially submerging the heat trace and flow transmitter.
The two ' A' division access pits, which are located adjacent to the flooded
'B' division pits, contained about an inch of water.
The lowest equipment
in these pits is about 6 inches from the bottom and was not covered.
The four pits form a row located on the south side of the emergency service
There is normally a concrete cover with a manhole over
water pump house.
each pit, but the covers were removed for installation of missile shields.
There were temporary covers of corrogated roofing material over the ' A'
division pits, but the covers on the flooded 'B' division had been removed.
There was also a plastic tent type structure over the entire area, but the
inspector found snow on the ground under the tent.
The inspector infonned the licensee of the problem and on February 2,1982
the pits were pumped dry with a portable pump and sand bags were placed on
the bottom inside walls of the tent type enclosure.
Licensee representatives
told the inspector that the cause for the flooding has not bean completely
determined, but that it appeared to be primarily due to surface run-off into
the top of the pit. The possibility of leakage around piping and conduit,
or through the walls, is still being evaluated.
. ..
. _ _ _
_
_
-
. - .
._.
. -_.
-
_
.
.
!
7.
!
System turn over from the constructive organization is accomplished separately
<
l
from the turnover of the facilities or areas in which the equipment is located.
The work on the missile shields was under the jurisdiction c' the Bechtel
construction organization, but the equipment inside the pit was turned over
to the licensee. Procedures do not adequately specify requirement for
- .
protecting turned over equipment within an area or facility from work activities
of the construction group. As a result, the pits were allowed to fill with
water to the extent that vital equipment was submerged.
Appendix B, Criterion II of 10 CFR 50 requires activities affecting quality
to be accomplished under suitably controlled conditions.
Controlled con-
ditions include suitable environmental conditions for accomplishing the
'
activi ty.
t
10 CFR 50, Aopendix B, Criterion V requires that activities affecting quality
shall be prescribed and accomplished by documented instructions appropriate
i
to the circumstances.
The Susquehanna PSAR, Appendix D, Section D.P.5 requires activities affecting
quality be prescribed by documented instructions, procedures or drawings
appropriate to the circumstances.
On February 5,1982 the inspector told the Superintendent of Plant that the
failure to have adequate procedures which establish sufficient environmental
controls to prevent flooding of safety-related equipment is a violation.
(387/82-04-03)
!
b.
Reactor Vessel Head
On February 11, 1982, the inspector witnessed the reactor vessel head being
put in place on the reactor vessel in Unit 1.
The inspector noted that the
work was being accomplished by an approved procedure, and that qualified
personnel were performing the job.
The inspector also visually inspected
selected head closure studs for both the reactor vessel head and the con-
tainment head.
i
No unacceptable items were noted.
.'
,
!
!
!
)
.
,
i
i
i
!
-
i
. _ - -
, . . . - . _ , -
.
. , - , . . _
, . _ - - . .
,- ,
.- . , - , , _ - .
8.
4.
IE Bulletin and Circular Followup
IE Bulletins and Circulars listed below were reviewed to verify the following:
(a) Bulletins and circulars received by PP&L corporate management were for-
warded to appropriate individuals with.n the organization, including
station management, for information, review and/or corrective actions as
required.
(b) PP&L bulletin responses were submitted to the fiRC within the specified
time period.
(c) Licensee reviews and evaluations of bulletins and circulars are complete
and accurate, as supported by other facility records and by inspector
observations of installed plant equipment.
(d) Corrective actions specified in licensee bulletin responses or internal
circular evaluation memoranda have been completed and/or responsibilities
have been assigned for completion.
Bulletin 80-21, " Valve Yokes Supplied by Malcolm Foundary Company, Inc."
--
The licensee replaced all of the defective valve yokes in Unit 1.
Six de-
fective yokes from Unit 2, and 2 yokes in spare valves still need replace-
ment.
These are being tracked by the licensee under Bechtel Noncompliance
Report 5980, and are scheduled for completion by October,1982.
This bulletin is closed.
-- Circular 80-02, " Nuclear Plant Staff Work Hours."
Administrative Procedure AD-QA-300, Revision 0, " Conduct of Operations" was
reviewed to detemine if concerns of the circular had been addressed.
Section
6.1.3 of the procedure complies with the guidelines established in the cir-
cular for licensed operators. However, it did not address these guidelines
for non-licensed operators or other personnel who perform safety-related
functions (e.g.: health physicists, I&C technicians, and key maintenance
personnel) .
This circular remains open pending resolution of this discrepancy.
.
d
-
.;
9.
Circular 80-18, "10 CFR 50.59 Safety Evaluations for Changes to Radioactive
--
Waste Treatment Systems."
This circular had been previously discussed in NRC Inspection Report 80-32
and was left open pending a review of a licensee procedure which addressed
proposed changes to the plant would require a safety evaluation.
On February
4,1982 Nuclear Department Instruction, NDI-QA-14.2.2, Revision 0, " Safety
Evaluations" was reviewed.
This procedure discusses which changes to the
facility must be evaluated and what the scope of the evaluation should be.
This circular is closed.
-- Circular 81-02, " Performance of NRC Licensed Individuals While On Duty."
Administrative Procedures AD-QA-300, Revision 0 " Conduct of Operations" and
AD-QA-303, Revision 0, "Shif t Routine" were reviewed. These two procedures
incorporated the concerns of the circular.
This circular is closed.
-- Bulletin 79-28, "NAMCO Model EA180 Limit Switches."
On February 17, 1981, Bechtel Nonconformance Report (NCR) 8162 was reviewed.
The NCR stated all Unit One NAMC0 switches applicable to the bulletin had
been repaired under PPSL Work Authorizations WA-U-12395 and WA-U-12399.
These work authorizations were reviewed and found to have performed the
su99ested repairs for all Unit One valves listed in the Bechtel NCR.
This bulletin is closed.
-- Bulletin 79-08, " Events Relative to Boiling Water Reactors Identified During
TMI Incident."
The licensee's commitments to NUREG 0737 and 0694 has been reviewed by NRC:
NRR and has been reviewed as documented in the Safety Evaluation Reonrt.
Future inspections to verify commitments made by the licensee base
1 NUREG
'
0737 and 0694 will be completed by Regional Inspections.
Since bulletin 79-08 information was essentially restated in one of the two
NUREGS, this bulletin is closed.
s
. .
_ _ . , - , , . .
,,,
,
- - - - -
. - , - - -
-
,
_ _
_
.
.
10.
Bulletin 79-09, " Failures of G.E. Type AK-2 Circuit Breaker in Safety
--
Related Systems.
This bulletin had previously been discussed in NRC Inspection Report
(387/81-25). On February 19, 1982 various AK-2 circuit breakers were in-
spected in the following 125 volt D.C. distribution panels:
-- 10-612
--
-- 10-642
No undervoltage devices as described in the bulletin were found in the
breakers.
This bulletin is closed.
l
I
!
!
l
If
.
.
-
11.
5
Preoperational Test Witnessing
Portions of the following preoperational testing were observed to verify that:
The approved test procedure was the current revision-being followed.
--
-- Qualified personnel were performing the test.
-- Test precautions and prerequisities were followed.
-- Test and measuring equipment met procedure requirements and was properly
calibrated.
Quality control hold and witness requirements were met.
--
Test results were properly documented.
--
-- Test procedures are technically adequate.
-- Test results are acceptable.
Criteria for interruption and continuation of testing are adhered to.
--
(a) Containment Atmosphere Control System
The approved procedure was reviewed prior to the scheduled testing on
January 27, 1982. The procedure did not verify the maximum design closure
time for containment vent and purge valves. The closing time requirements
are listed in FSAR Table 6.2-12, and in Technical Specification Table
3.6.3-1.
The inspector asked the Test Engineer if the valve closing times
were measured as part of a different procedure and found that they were
not. The inspector told the Test Engineer that containment isolation valve
design closing times must be verified, and that the preoperational test
procedure appeared to be inadequate.
On January 28, 1982 the inspectors looked at the rest of the preoperational
test procedures involving containment isolation valves to see if closing
times were measured.
Containment isolation valve closing times were not
measured in preoperational tests P25.1, Primary Containment Instrument
Gas, and P34.2, Reactor Building Chill Water.
In test P64.1, Reactor Cir-
culation System, the valves were timed, but no acceptance criterion was
listed. Test P50.1, Reactor Core Isolation Cooling, required valve 1F088
to close in less than 5 seconds, however, FSAR Table 6.2-12 requires a 3
second closing time.
Valve 1F088 actually closed well within the 3 second
'
limit during testing.
In test PS2.1, High Pressure Coolant Injection, valves
IF100 and IF042 were not timed.
-
-
(11.
__
_
.
.
12.
Appendix B, Criterion XI of 10 CFR 50 requires preoperational testing of
structures, systems and components to demonstrate that they will perform
satisfactorily in service. The tests must be performed in accordance with
written test procedures which incorporate the requirements and acceptance
limits contained in applicable design documents.
Quality Assurance Procedure SP-3, Control of Testing and Inspection Activities,
Part 5.4.12 requires the preoperational test program to assure that licensing
commi t i. :nt - and design specifications are reflected in the completed installa-
tions.
The Se quehanna FSAR, Section 14.2.12.1 (P59.1) states that the closure
times specified in the FSAR for containment isolation valves are to be veri-
fied in the various system preoperational tests.
Failure to properly verify closure times of the containment isolation valves
identified above is a violation.
(387/82-04-04)
This violation was discussed with the Assistant Station Superintendent on
January 28, 1982. He informed the inspectors that the Integrated Startup
Group (ISG) had also initiated a review of other preoperational tests to
determine the extent of the problem. On January 29, 1982, the Assistant
Station Superintendent informed the inspectors that a new preoperational
test would be issued for the purpose of testing containment isolation valve
closure times, and that the appropriate change to the FSAR would be submitted.
The inspectors reviewed the draft Technical Specifications for the Susque-
hanna Steam Electric Station, Unit 1.
Section 3.6.3 states limiting condi-
tion for operation for primary containment isolation valves, and states
that the isolation valves listed in table 3.6.3-1 must be demonstrated oper-
able with isolation time less than or equal to those listed on table 3.6.3-1.
Since the valve timing would have to be performed to meet this requirement
prior to entering operational condition 3 (Hot Shutdown) the violation for
not performing time testing in the preoperational tests is not considered
as having large safety significance.
On February 4,1982 preoperational testing was observed. No unacceptable
items were identified.
b.
On the evening of February 10, 1982 portions of Section 8.3.1 of Preopera-
tional Test P55.1, Revision 2 were witnessed.
No discrepancies or unacceptable items were noted.
1
-
- _ ~ .
_
, - - -
. . ,
_ _ _ _ _ _ _ _ _
l
l
13.
6.
Preoperational Implementation
a.
I
I
On February 5, 1982 ' two position switch was observed connected to ter-
minals 1 and 2 in terminal box TB0144 in the Reactor Core Isolation System
(RCIC) Room. The switch had no temporary modification tag attached to it.
The temporary modification log was reviewed, but no entry could be found for
putting this switch into the terminal box.
The Integrated Startup Test Engineer for the RCIC System stated the switch
was placed in the system to be used as a local safety trip during initial
testing.
10 CFR 50, Appendix B, Criterion V states that activities affecting quality
shall be prescribed by approved procedures.
PP&L Quality Assurance Manual Procedure SP-9, Revision 1, Section 5.3 states
that temporary modifications shall be controlled and documented in accordance
with approved procedures.
PP&L Startup Administrative Manual Procedure AD6.8, Revision 4, Sections 5.3
a'id 5.4 states that an orange tag is used for field identification of temporary
modifications, and that temporary modifications are documented on the Temporary
Modification Log.
On February 5,1982 the Superintendent of Plant was notified that since the
switch had not been entered into the temporary modification log, nor had a
orange temporary modification tag been placed on the switch, this was a
violation of 10 CFR 50, Appendix B, Criterion V.
(387/82-04-05)
b.
Controlled Drawings
On February 12, 1982 controlled drawing stick file number 87 and 127 were
reviewed.
It was noted that stick file 127 drawing E-137, sheet 9 was re-
vision 2 of the drawing while stick file 87 had revision 3 to the drawing.
On February 16, 1982 the inspector reviewed the Audit Verification Sheet for
stick file 127. The audit was begun on January 30, 1982 and completed on
February 15, 1982. The audit verification sheet noted that drawing E-137
sheet 9 was the wrong revision, and en updated version had been requested.
No unacceptable items were identified.
t
T
.
h
14.
7.
Startup Test Program Review
The inspector reviewed the licensee startup test program against the commit-
,
ment in FSAR and the requirements of Regulatory Guide 1.68, Revision 1.
Based
on the above review, the inspector had several questions on how the licensee
complies with the requirements of certain paragraphs of R.G.1.68, Revision 1,
Appendix A and discussed them with the licensee personnel.
Some of these ques-
tions were satisfactorily resolved by the detailed explanation presented by
the licensee personnel. The remaining questions, described below, need further
actions by the licensee for their resolution:
R.G.1.68, Revision 1, Appendix A, Paragraph 5.n requires collection of
baseline data for reactor coolant system loose parts monitoring system
during the power ascention test phase. This test is not included in any
of the Startup Test Procedure abstracts in FSAR Section 14.2.12. 2. The
licensee personnel stated that this test will be included as a subtest of
one of the presently scheduled startup tests.
This is an insnector followup item.
(387/82-04-06)
R.G.1.68, Revision 1, Appendix A, Paragraph 5.e.e requires demonstration
of the primary containment inerting and purge system operation in accordance
with design, during the power ascension test phase. This test is not in-
cluded in any of the startup test procedure abstracts in FSAR Section 14.2.
12.2.
The licensee personnel stated that a new subtest procedure (ST 37.2)
will be written and incorporated into an existing test procedure (ST 37,
Gaseous Radwaste System) and the new subtest will be scheduled to be per-
formed following the Plant Warranty Run.
This is an inspector followup item.
(387/82-04-07)
R.G.1.68, Revision 1, Appendix A, Sections 4 and 5 require low power
tests and power ascention tests to be performed at definite power plateaus
( 5%, 25%, 50%, 75%, 100%) or at definite power ranges,
The startup test
program as described in FSAR Section 14.2 indicates that these tests to
be performed at certain Test Conditions (TC-0,H,1,2,3,4,5, and 6). These
test conditions generally do not correspond to definite power plateaus
required by R.G.1.68, but vary over wide range of power levels. This is
especially true for TC-1,2, and 3.
The licensee personnel explained that a
new startup test procedure (ST-99) will be written to specify various power
plateaus or ranges for each test as required by RG 1.68.
This is an inspector followup item.
(387/82-04-08)
-
1
-
.
15.
8.
Startup Test Procedure Review
The following preliminary issues of startup test procedures were reviewed for
compliance with NRC requirements and licensee commitments:
a.
ST-4, Revision 0, Full Core Shutf7wn Margin.
b.
ST-15, Revision 0, High Pressure Coolant Injection.
c.
ST-25, Revision 0, Main Steam Isolation Valves,
d.
ST-27, Revision 0, Turbine Trip and Generator Load Rejection,
e.
ST-28, Revision 0, Shutdown From Outside The Control Room.
Based on the above review, the inspector had several questions 2nd discussed
them with licensee personnel. These questions were satisfactorily resolved
by the answers presented by the licensee personnel which were further verified
by the inspector by review of pertinent records except that one question needs
further licensee action for its resolution. The following is a brief outline
of these questions and their resolution:
a.
ST-15, Revision 0, High Pressure Coolant Injection
FSAR Section 14.".7 commits to R.G.1.68, Revision 1, which requires
demonstration of auto start of HPCI system under simulated accident
'
conditions and injection into the reactor coolant system during power
ascension test phase (see R.G.1.68, Appendix A, Paragraph 5.k).
The
'
inspector questioned the validity of simulating the accident conditions
by using the MANUAL INITIATION push button rather than by tripping the
accident condition primary sensors (reactor low level switches and dry-
well high pressure switches). The licensee stated that the circuitry
for HPCI auto initiation from the primary sensors is in parallel with
the MANUAL INITIATION push button and will be tested during preoperation
testing, and therefore is not repeated during startup testing. The
inspector verified this by review of HPCI Preoperational Test P52.1, Revision 2.
.
l
.
!
,
'
!
-
.
16
b.
ST-25, Revision 0, Main Steam Isolation Valves
The insoector pointed out that R.G.1.6R Revisinn 1 Anoendix A Para-
graphs 4.m
5.u, and 5.m.m require testin,g of MSIV's,at less than, 5%
power (Low Power Test) and at 25%, 50%, and 100% po;.er (Power Ascension
s.
Tests) and that the tests as described in ST-25 generally satisfy these
requirements except that the Power Ascension Tests at 25% and 50% pov:er
are grouped as a single test to be performed between 5% and 75%. The
licensee personnel explained that there will be two separate tests
(near 25% and 50% power) as shown on the draft Power Ascention Test Pro-
gram Schedule. The inspector verified this by review of the above
schedule.
c.
ST-27, Revision 0, Turbine Trip and Generator Load Rejection
The inspector pointed out that Section 27.2.2, Initial Status, of ST-27
dots not specify the Recirculation Flow Control Systems mode (AUT0/
MASTER MANUAL) during the performance of this test and the selection of
the mode will influence the test results.
The licensee personnel ex-
plained that the MASTER MANUAL mode will be selected as shown on FSAR
Figure 14.2.5, Sheet 1, and will be specified in G0-00-003, Operating
Procedure for Power Ascension, Revision D, which will be used to arrive
at the initial conditions for ST-27. The inspector verified this by re-
view of the above operating procedure,
d.
ST-28, Revision 0, Shutdown From Outside the Control Room
The inspector pointed out that R.G.1.68.2 (Initial Startup Test Program
to Demonstrate Remote Shutdown Capability for Water-Cooled Nuclear
Power Plants), Revision 1, which the licensee is committed to, requires
maintaining stable hot standby condition for at least 30 minutes during
this test and that the test as described in ST-28 does not reflect this
requirement.
The licensee personnel explained that the decay heat from
the fresh reactor core is not sufficient to overcome the losses from
the reactor coolant system and, therefore, the above requirement cannot
be achieved in practice.
The inspector stated that the test abstract in FSAR Section 14.2.12.2
for ST-28 describes the initiation of this test with a reactor scram
from outside the Control Room where as the test as described'in the
detailed test procedure (ST-28, Revision 0) describes the initiation of
the test with a reactor scram from the Control Room. The licensee
personnel explained that FSAR Section 14.2.12.2 will be corrected to
agree with ST'-28, Revision 0.
This is an inspector followup item.
(387/82-04-09)
_
_
.
.
,
17.
9.
Quality Assurance Survbillance Activities
Operational Policy Statement (OPS) Number 7, Revision 0, states in Section
5.3 that Nuclear Quality Ass' rance (NQA) will establish a Surveillance Program
L
u
to verify by observation that activities are performed in specified manners.
It also states that Functional Unit Procedures will be established to detail
and implement this surveillance program.
On February 25, 1982 the Operations Quality Assurance Supervisor was questioned
as to status of functional unit procedures for the surveillance program. He
stated a Surveillance Program Procedure was in draft
and being reviewed.
,
The Operational Quality Assurance Program must in effect at least 90 days
'
prior to receiving an operating license; and therefore the Surveillance Pro-
gram mutt be in effect by that time. The approved program will be reviewed
during a subsequent inspection.
(387/81-04-10)
10.
Emergency Planning Meeting
On the arening of February 24, 1982, at the request of the Secretary of the
Nescopeck Borough Council the inspector and the Director of the Division of
Emergency Plans and Operational Support attended a meeting organized by the
Berwick Corough Council with other Columbia County Municipalities and town-
ships. The meeting was organized to listen to a presentation made by repre-
,
~
sentatives of the licensee on Emergency Planning.
11.
Traversing In-Core Probe System Electrical Power Supplies
On January 27, 1982 representatives from PP&L told the NRC Resident Inspector
that power supplies to the traversing In-Core Probe System (TIPS) contain-
ment isolation ball valve, and to tne back-up shear valve were not installed
as safety-related. Since the ball valve is considered the containment isolation
valve, a PP&L Quality Assurance Action Request (Number 82-011) had been initiated
on January 26, 1982 requesting the PP&L Nuclear Plant Engineering (NPE) Group
to evaluate this problem. On January 27, 1982 the NPE Group responded stating
that the power supplies to both the ball valve and the shear valve should be
safety grade.
PP&L Nonconformance Report Number 82-058 was written to document
the problem.
The problem was then discussed with the General Electric (G.E.) Operations
.
Manager at the Susquehanna Site who stated that all G.E. BWR's would have the
same electrical set-up since G.E. did not consider the TIPS System as safety-
related.
i
FSAR Table 3.2-1 states that TIPS piping and isolation valves are Safety Class
f
2 and Seismic Category I, but does not discuss the power supplies for the valves.
With the present configuration, it can not be assured that the ball valve or
the shear valve will be capable of operation when a containment isolation signal
is generated.
The resolution to this problem will be further addressed by the NRC and will be
reviewed during a subsequent inspection.
(387/82-04-11)
l
t
i'
,
.
_
.
r
1
.
.
1
18.
12.
Exit Interviews
At periodic intervals during the course of this inspection, meetings were held
with facility management to discuss the inspection and findings identified.
,
m
!