ML20154M025
| ML20154M025 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 10/15/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20154L952 | List: |
| References | |
| 50-289-96-201, 50-289-98-06, 50-289-98-6, NUDOCS 9810200142 | |
| Download: ML20154M025 (39) | |
See also: IR 05000289/1998006
Text
- i
U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No. 50-289
License No. DPR-50
Report No. 98-06 -
r
Licensee: GPU Nuclear Corporation
Facility: Three Mile Island Station, Unit 1
!
Location: P.O. Box 480
Middletown, PA 17057
Dates: August 10 through August 27,1998
Inspectors: J. Yerokun, Senior Reactor Engineer
R. Bhatia, Rear: tor Engineer
C. Baron, NRC Contractor
i
!
Approved by:
'
Eugene Kelly, Chief
Systems Engineering Branch
Division of Reactor Safety
i
F
F
!
i
9810200142 981015 P "
PDR ADOCK 05000289 >
G PM ,
. ., .
.
.
, f ,
. .
i
EXECUTIVE SUMMARY ,
';
Three Mile Island Nuclear Power Station
Report No. 50-289/98-06 t
August 10,1998 - August 27,1998 ;
!
The inspection examined engineering activities associated with findings from the NRC
Architect Engineering (AE) Team inspection that was completed in January.of 1997 and '
,
documented in NRC Inspection Report 50-289/96-201. The inspection also examined the
effectiveness of the corrective actions taken to address the violations stemming from the ' '
issues identified by the AE team and documented in the Notice of Violation and Proposed
'
imposition of Civil Penalties dated October 8,1997. >
e Although several iterations were required, engineering personnel addressed the !
issue of a letdown line break outside containment and verified that safety related 1
equipment in the auxiliary building, would remain operable following a letdown line
- break. (E8.1)
-o- The concerns with the high pressure injection system involving coordination of the
opening of the suction and discharge valves with pump MU-P-1C operation; ;
potential for gas accumulation in the pump suction piping; and the stroke times for ;
system valves specified in surveillance procedure 13OO-3H,"lST of Makeup Pumps ,
and Valves," were resolved. ~(E8.2, E8.3, E8.5) ;
e A violation of NRC requirements was identified with the changen made to the . !
position of the cross connect valves. The safety evaluation failed to identify that -
the change involved an unreviewed safety question, as the change introduced a
malfunction of a different type than any evaluated in the UFSAR;1.e., a failure of
the "C" HPl pump due to gas entrainment. (E8.4)
e Engineering personnel adequately demonstrated that during a seismic event, the
auxiliary steam piping will not fail. No physical changes were required to arrive at
]
this conclusion and the issue did not represent any past operational concern. (E8.6) j
e The issues associated with makeup pumps' net positive suction head (NPSH) were
'
resolved, if the makeup tank is depressurized, or if there is a low-low level in the
borated water storage tank, adequate NPSH will still be available for the pumps.
(E8.7, E8.8)
e Voltage drop analysis and calculations showed that safety related equipment,
including MU-V-18, would have sufficient voltage to perform their design function.
(E8.9)
e Operation of makeup system valve MU-V-18 prior to January 8,1997, was not
appropriately controlled by procedures. Because it did not represent a more broad
concern with respect to establishment of procedures for safety activities, this minor
violation was not subject to formal enforcement action. (E8.10)
o
ii
_ . . _ . . _ _ _ > _ . . _ _ _ . _ . _ . . . _ _ -.___ _. _ _ _ ._ _ .. _ _-. _ . _ .. _ _ _ _
i i
e. Inconsistencies in data sheets of calibration surveillance procedures, such as
instrument uncertainties, static head corrections and instrument loop errors were
adequately resolved. (E8.11, E8.21, and E8.22)
e FSAR and technical specification discrepancies identified by the AE team were
properly resolved and the ongoing efforts by the licensee to improve the overall
quality of the FSAR appeared appropriate. (E8.12, E8.15, E8.16, and E8.26)
- Adequate corrective actions were taken to address the issues associated with
potential air entrainment into the decay heat removal system pump suction during
the switch from the Borated Water Storage Tank to the Reactor Building sump.
(E8.13)
!
e The corrective actions taken to address the violation stemming frcm improperly
taking credit for containment overpressure in low pressure injection (LPI) pumps
NPSH calculations were appropriate. New calculations reflected that adequate
NPSH would be available for the LPI pumps without taking credit for containment
overpressure. (E8.14)
e Programmatic issues such as inservice testing program discrepancies involving
DHRS valves, and Corrective Action Program deficiencies with tracking open items
associated with safety system functional inspection (SSFI) and service water
system operational performance inspection (SWSOPI) activities were resolved.
(E8.17a,17b,17c, and E.18)
! e Engineering evaluation confirmed that there was no safety concern associated with
reverse flow through the reactor building (RB) sump screens. (E8.19)
l; e The discrepancy associated with the use of non safety-related annunciation to
accomplish safety related actions by operators was appropriately resolved. Good
procedural guidance and training were provided. (E8.23)
e The actions taken to correct the improper attachment of a hose and metal tubing to '
the discharge of relief valve BS-V-63B were adequate. This minor violation was
not subject to formal enforcement action. (E8.24)
e Appropriate actions were implemented to ensure that the BWST level transmitter
, enclosure and heat tracing were maintained properly. (E8.25)
t
1
e The actions taken and those ongoing to identify and control calculations / analysis
were adequate. Enhanced procedure, EP-OO6, " Calculations," was comprehensive
l and contained good guidance for controlling calculations. (E8.28)
i
,
!
'
iii
. _ _ _ _ . _ . _ _ _ _ _ ._ _ _ . _ .. _.-. _ _ _ .. _ _ _ . _ _._.. _ . _ _ _
i .
-
.
'
TABLE OF CONTENTS
i
PAGE ;
EX ECUTIV E S U M M ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . il
111. Engine ering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
E8 Miscellaneous Engineering issues (92903) ...........................1 ,
E8.1 ' (CLOSED) IFl 50-289/96-201-01, Letdown line Break in '
the Auxiliary Building . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
E8.2 (CLOSED) IFl 50-289/96-201-02, Evaluation of Simultaneous ,
Start of MU&P Pump MU-P-1C and Suction and Discharge Valves . . . . . . 2 (
E8.3 (CLOSED) IFl 50-289/96-201-03, Evaluation of Gas
c Accumulation in Suction Piping for Makeup Pump MU-P-1C . . . . . . . . . 4 :
E8.4 (CLOSED URI 96-201-04, and eel 97-070-01023, Adequacy of Makeup ,
Tank Pressure / Level Curves ................................5 !
E8.5 (CLOSED) URI 50-289/96-201-05, Design Basis Valve Stroke Times in
Surveillance Procedure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
E8.6 (CLOSED) IFl 50-289/96-201-06, Consequences of Failure of Auxiliary
St e a m Piping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 -
i
E8.7 (CLOSED) IFl 50-289/96-201-07, Loss of Pressure in MU&P Tank
due to Letdown Line Break . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
E8.8 (CLOSED) IFl 50-289/96-201-08,M&UP Pump NPSH When
Taking Suction From BWST or Makeup Tank . . . . . . . . . . . . . . . . . . . . 10
E8.9 (CLOSED) IFl 50-289/96-201-09, Incomplete DC System Voltage Drop
i-
Calculations ..........................................11'
E8.10 (CLOSED) URI 50-289/96-201-10, Alignment of MU-V-18 DC
- Po w e r S u p ply . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 i
E8.11 (CLOSED) URI 50-289/96-201-11, Makeup Tank Level Instrument
Loop Tolsrances .......................................13
'
E8.12 (CLOSED) URI 50-289/96-201-12,FSAR Discrepancies . . . . . . . . . . . . 14
4
E8.13 (CLOSED) URI 50-289/96-201-13,and eel 97-070-01013, i
Adequacy of BWST Setpoint for DHRS Pump Switch over to !
4 RBSump............................................14
, E8.14 (CLOSED) URI 50-289/96-201-14,and eel 97-070-01093,
Adequacy of Safety fivaluation for an FSAR Change Crediting
Containment Overpmssure for LPI Pumps NPSH . . . . . . . . . . . . . . . . . 16
!
E8.15 (CLOSED) IFl 50-289/96-201-15, Technical Specification :
i Discrepancy ...... ...................................17
'
E8.16 (CLOSED) URI 50-289/36-201-16, Discrepancy Between .
i FSAR and Technical Specifications Regarding DHRS Leakage . . . . . . . . 18
- E8.17a(CLOSED) URI 50-289/96-201-17A,and eel 97-070-10143,
- Leak Testing of DHRS Pump Suction Check Valves . . . . . . . . . . . . . . . 19
, E8.17b(CLOSED) URI 50-289/96-201-178,and eel 97-070-08014, i
i Leak Testing of DHRS Pump Discharge Check Valves .............20
1
a
1 iv l
?
a
.
'l
5
- _ _ _ . . _ _ ._ - _ . . . _ _
- ._ . - . - -.. .. . - - - - - . . .- . - . . . . - - -
i
.
TABLE OF CONTENTS (CONT'D)
PAGE
E8.17c(CLOSED) URI 50-289/96-201-17C,and eel 97-070-08014,
inspection of DHRS Pump Vault Floor Drain Check Valves . . . . . . . . . . 21
E8.18 (CLOSED) URI 50-289/96-201-18,and eel 97-070-03013,
Timeliness of Action on SSFI Open items . . . . . . . . . . . . . . . . . . . . . . 22
E8.19 (CLOSED)IFl 50-289/96-201-19, Evaluation of Reactor Building Sump
Screen for Reverse Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22
E8.20 (CLOSED) URI 50-289/96-201-20,and eel 97-070-07014, Testing of
Molded Case Circuit Breakers . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23
E8.21 (CLOSED) URI 50-289/96-201-21,and eel 97-070-01033, Static Head
Correction for BWST Level Transmitter . . . . . . . . . . . . . . . . . . . . . . . . 24
E8.22 (CLOSED)IFl 50-289/96-201-22,BWST Level Instrument Drift . . . . . . . 25
E8.23 (Closed) URI 50-289/96-201-23, Selection of BWST Low Level
Alarm f or Operator Action . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26
E8.24 (CLOSED)IFl 50-289/96-201-24, Undocumented Modification . . . . . . . 27
E8.25 (CLOSED) URI 50-289/96-201-25,and eel 97-070-06014, BWST Level
Transmitter Enclosure and Heat Tracing . . . . . . . . . . . . . . . . . . . . . . . 28
E8.26 (CLOSED) URI 50-289/96-201-26,FSAR Discrepancies . . . . . . . . . . . . 29
E8.27 (CLOSED)IFl 50-289/96-201-27, Definition of Single Active Failure . . . 29
E8.28 (CLOSED) URI 50-289/96-201-28,and eel 97-070-10153,
10163,10173, and 0183, Control of Calculations . . . . . . . . . . . . . . . . 30
V. M anagem ent Meeting s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31
X1 Exit Me eting Summ ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31
v
. . - - - . . . - - - - - . - - . - . - . - . - - . . .- . _ . .- - ..
1 .
i
l -
l'
i Report Details
111. Enaineerina
E8 Miscellaneous Engineering issues (92903)
Between November 11,1996 and January 10,1997, an NRC Architect Engineering
(AE) Team conducted a design inspection at TMI-1. The results were documented
- in NRC Inspection Report 50 289/96-201. The team identified twenty eight items
i _ as either Unresolved or inspector Follow up items (URIs or IFis). Some of the items
t
were later identified by the NRC as violations of NRC requirements and transmitted
!
' to GPU Nuclear in a letter dated October 8,1997. This inspection was conducted
to review the resolution of the remaining open items and the corrective actions
,
taken to address those already identified by the NRC to be violations. The results of
I
these reviews are documented below.
E8.1 (CLOSED) IFl 50-289/96-201-01. Letdown line Break in the Auxiliarv Buildina
a. Incoection Scone
The AE team identified a discrepancy involving the licensee's consideration of the
! effects of a letdown line break in the auxiliary building. It was not clear to what
l extent the licensee was required to consider the effects of a letdown line break in
l the auxiliary building. i
i !
l b. Observations and Findinas
i
'
l
'
At TMI-1, high energy lines are defined as piping containing fluids that exceed both
200*F Endr 275 psig. The reactor coolant system (RCS) letdown line in the auxiliary
building normally operates below 200*F and therefore, based on the existing
licensing basis definition, it appeared that full diameter breaks did not have to be l
postulated, although the pressure exceeds 275 psig. However, the licensing basis
l for TMI-1 included the licensee's response to the NRC's request for additional
information (RAl) involving the licensee's evaluation of six full diameter letdown line j
break locations described in the FSAR, Appendix 14A. Therefore the AE Team '
determined that the TMi-1 licensing basis for pipe breaks included the postulation of
full diameter letdown line breaks between the containment penetration and the
breakdown orifice as described in Appendix 14A of the FSAR. Consequently, the
design of safety related equipment in the affected areas should consider the
,
conditions resulting from these breaks.
!
i.
L Following further reviews, GPU Nuclear agreed that high energy line break (HELB)
'
consideration for the letdown line outside containment was part of the licensing ;
basis for TMI-1 and therefore initiated actions to complete a break analysis for the
line.
!
!
$ l
L
.
4
2
The inspector reviewed the licensee's operability determination (PRG No.97-072
and 98-22 meeting minutes) asserting that equipment in the pipe area would not be
adversely affectea by a line break. The inspector identified weaknesses and errors
in the determination. For example, the bases for equipment operability were
unclear. The licensee entered this issue into their corrective action program (CAP
No. T198-0666)to address the inspector's concerns. The issue was reviewed
further by the licensee and a new operability determination was documented in PRG
Meeting No. 98-76 (August 24,1998). Technical Data Report (TDR) No.1230
documented the evaluations and engineering judgement made regarding the
capability of TMI-1 auxiliary building safety-related equipment to remain functional
when subjected to a postulated letdown line break environment. Since the letdown
line break was not part of the design bases, the licensee did not include the affect
safety related equipment in their Environmental Qualification (EO) list. The
inspector did not identify any discrepancy with this assertion.
The inspector reviewed operability determination documented in PRG Meeting
Minutes No. 98-76 and TDR No.1230. The licensee stated that the design basis
of electrical equipment at TMI-1 is IEEE Report NSG/TCS/SC4-1 of 1969. Based on
the expected service condition from a letdown line break, electrical equipment
would remain operable in accordance with the IEEE report. Valves and cables were
determined to be qualified to more severe conditions than would be expected from a
letdown break. Since the licensee could not locate an original analysis of a break in
the letdown line of the auxiliary building, the analysis contained in the TDR
represented GPU Nuclear's analysis of record for letJown line break in the auxiliary
building. This analysis therefore served as the basis for the statements in
Amendment 41 to the operating license and now contained in the FSAR. The
inspector found no discrepancy and determined that the TDR properly addressed the
issue. The inspector conducted a walkdown of the area around the letdown line in
the auxiliary building and verified that the safety-related equipment in the area were
addressed in the TDR.
c. Conclusions
Although several iterations were required, engineering personnel addressed the
issue of a letdown line break outside containment and verified that safety related
equipment in the auxiliary building, would remain operable following a postulated
letdown line break.
E8.2 (CLOSED) IFl 50-289/96-201-02. Evaluation of Simultaneous Start of MU&P Pumo
MU-P-1C and Suction and Discharae Valves
a. Inspection Scoce
The AE team identified a discrepancy involving make up pump MU-P-1C and the
opening stroke times of the suction and injection valves. The specific concern was
over the effect on the pump of a slow opening suction valve and a fast opening
high pressure injection valve.
-
_ _. -
_
i
. ;
i
i
3 ;
b. Observations and Findinas !
,
The suction line of pump MU-P-1C was normally isolated from makeup tank MU-T-1
by valves MU-V69A and MU-V69B. It is also isolated from the BWST by motor
operated valve MU-V-14B. In the event of an accident the pump and the motor i
operated valve receive an actuation signal at the same time. No interlocks or time
delays were provided to start the pump until after the suction valve had opened.
MU-P-1C would not be aligned to any water source until MU-V-148 automatically
opened in response to an actuation signal.
- !
The licensee evaluated this issue and determined that pump MU-P-1C was operable. )
- This determination was based on a 1974 preoperational test results that included
,
the simultaneous start of pump MU-P-1C and opening of valve MU-V-148, and on a I
review of the test records for valve MU-V-148. During the preoperational test, MU- .
V-148 stroke time was nine seconds. The licensee stated that the opening stroke l
time of valve MU V-148 had remained at nine seconds or less, based on a review of l
'
test records. Since the pump achieved full speed within six seconds, the openmg i
time for the suction valve was adequate to ensure flow to the pump. The high l
l
pressure injection valve in the pump discharge line was estimated to open fully in '
ten seconds. However, the maximum design basis stroke time for MU-V-14B
opening is 22 seconds in Surveillance Procedure (SP) 13OO-3H,"lST of MU Pumps ,
and Valves."
The licensee determined that the reference stroke time of 13 seconds in SP 1300-
3H would be adequate to ensure the required makeup pump net positive suction
head (NPSH). Therefore, SP 1300-3H was revised to reduce the design basis
opening stroke time for both MU V-14A and MU-V-148 from 22 saconds to 13
seconds, in addition, the normal positions of makeup pump suctirn cross connect
manual valves MU-V69A and MU-V69B were changed from closed to open. With
these valves open makeup pump MU-P-1C suction would no longer be isolated from
the makeup tank MU-T-1 during pump startup. However, a seprirate concern
created by the changing of these valves' position is addressed in conjunction with j
Unresolved item 50-289/96-201-04," Adequacy of Makeup Tank Pressure / Level -
Curves."
The inspector determined that the specified stroke time of 13 seconds was
appropriate for the suction valves. The inspector reviewed revision 46 of
surveillance procedure (SP) 1300-3H, "lST of MU Pumps and Valves," and verified
that the design basis stroke times for MU-V-14A and MU-V-14B opening was 13
seconds. Based on the licensee's evaluation, this change would ensure that the
required makeup pump NPSH would be available under accident conditions. The
inspector also reviewed drawing 302-661," Makeup & Purification System Flow
Diagram," Revision 50 and Operating Procedure 1104-2," Makeup and Purification
System," Revision 113 to verify that makeup pump suction cross connect manual
valves MU-V69A and MU-V69B were normally in the open position.
l
'. g 1
4
. c. Conclusions
The licensee ensured that upon automatic actuation of the HPl System, the opening
of the suction and discharge valves would be in sufficient time to not create any .
adverse effect on pump MU-P-1C operation. ;
.E8.3 (CLOSED) IFl 50-289/96-201-03, Evaluation of Gas Accumulation in Suction Pinina
for Makeuo Pumo MU-P-1C
a. Inspection Scone
The AE team identified a discrepancy involving the potential for accumulation of
non-condensibles, such as hydrogen, released from the stagnant water in the
suction line of makeup pump MU-P-1C because of the physical configuration of the
pump's suction piping.
' b. Observations and Findinas
The pump's suction was normally isolated from makeup tank MU-T-1 by valves MU.
V69A and MU V698. It is also isolated from the BWST by motor operated valve
MU-V-14B. The isolated piping section was approximately 140 feet long with a
long horizontal section approximately 10 feet above the makeup pump suction
elevation. Other than quarterly cycling of motor operated valve MU-V-14B and
ensuring that the piping was filled and vented after any maintenance work that
required draining of the piping, no monitoring for gas accumulation was performed.
The licensee instituted a daily check of Pump MU-P-1C suction pressure as an
interim measure during the design inspection. The licensee found that if the
. pressure was maintained at greater.than 30 psig, there was no mechanism that
- could have caused the accumulation of non-condensibles in the piping, in addition
to this issue, the licensee identified a condition where the isolated suction piping for
the pump may have been overpressurized during testing and could similarly have
been during the system response to a design basis accident. This condition was
documented in Licensee Event Report (LER) 97-0031," Potential Over
pressurization of Makeup Pump Suction Piping due to inadequate Test and
Operating Procedures," dated September 29,1997. The LER concluded that there
were no adverse safety consequences or safety implications that resulted from the
condition, and that the event did not affect the health and safety of the public.
In response to these issues, the licensee changed the position of the makeup pump
. suction cross connect valves MU-V-69A and MU-V-69B from normally closed to
open. Safety evaluation number 000211-015," Operation with MU Suction X-
connect Valves Open," was issued to support this change. The revised valves
positions were shown on drawing 302-661," Makeup & Purification Flow Diagram,"
Revision 50. As a result of this change, the suction of all three makeup pumps are
normally aligned to the makeup tank.
. _ . . _ . . _. _ _ . _ . _ - - _.- ..~ . _ . _ - _ . _ _ - . . . _ _ _ _ . ___
. ..
t
5
The inspector reviewed the licensee's evaluations and discussed the issue with
engineering personnel. The inspector determined that opening the pump suction
valves resolved the gas accumulation concern. The inspector reviewed drawing
.302-661," Makeup & Purification System Flow Diagram," Revision 50 and
Operating Procedure 1104-2," Makeup and Purification System," Revision 113 and .
verified that makeup pump suction cross connect valves MU-V69A and MU-V698
were normally in the open position. However, a separate concern created by the
-
changing of these valves' position is addressed in conjunction with Unresolved item
50-289/96-201-04," Adequacy of Makeup Tank Pressure / Level Curves."
c. Conclusions
The concern witn gas accumulation in makeup pump MU-P-1C suction piping was
resolved. . The monitoring of the makeup pump suction pressure when the concern
was raised was appropriate and changing the position of the pumps' suction cross i
connect valves resolved the problem.
E8.4 (CLOSED URI 96-201-04, and eel 97-070-01023. Adeauacy of Makeuo Tank
Pressure / Level Curves
a. Insoection Scope
The AE team identified a concern associated with the pressure and level operating
limits of makeup tank MU-T-1 in calculation C-1101-211-5310-0047," Makeup
. Tank Drawdown during LOCA." Specifically, there were inconsistencies in the
'
calculation involving the makeup tank volume / height relationship, omission of some
piping and fittings, use of non-con'servative pump flowrates, failure to account for
vortexing in the makeup tank,'and failure to consider the conservative minimum
BWST level. Subsequent NRC reviews determined that the discrepancies
constituted a violation (eel 97-070-01023)of 10 CFR 50, Appendix B, Criterion lil,
" Design Control," as documented in a Notice of Violation and Proposed imposition
of Civil Penalties, dated October 8,1997.
b. Observations and Findinas
To correct this problem, the licensee performed new calculations, and revised
applicable procedures and the FSAR. Calculation C-1101-211-E610-066," Makeup
Tank Level and Pressure Limits," was performed to account for a lower back
pressure, procedural limitations to flow, and to correct the other non-conservative
assumptions identified in the violation. Calculation C-1101-211-E540-085,"TMI-1
Makeup Tank Gas Entrainment Determination," addressed potential vortex formation
in the makeup tank. Operating Procedures 1101-1," Plant Limits and Precautions,"
and 1104-2," Makeup and Purification System," and Abnormal Transient Procedure
1210-01," Reactor Trip," were revised to be consistent with the new analyses.
Section 6.1.3.1 of the FW was revised to reflect the results of the analyses.
Safety Evaluation Number 000211-019,"MU Pump NPSH and Gas Binding
Concerns Evaluation," was issued to support this FSAR change.
, _ --- __.
i
.. .,
'
l
6
)
The inspector reviewed calculations C-1101-211-E610-066," Makeup Tank Level
and Pressure Limits," Revision 1: C-1101-211-E610-068," Makeup System Model
for MU Pump NPSH Calculations," Revision 0; and C-1101-211-E540-085,"TMI-1 ]
Makeup Tank Gas Entrainment Determination," Revision 0 and verified that the '
issues identified during the design inspection had been resolved. The inspector also
verified that the makeup tank pressure / level limits included in Figure 1 of Operating I
Procedure 1104-2, " Makeup and Purification System," Revision 113 were 1
consistent with the results of the calculations, and that makeup and purification
system information had been deleted from Operating Procedure 1101-1, " Plant l
Limits and Precautions," Revision 64. In addition, the inspector verified that
Operating Procedure 1104-2," Makeup and Purification System," required normal )
makeup isolation valve MU-V-222 to be locked in the throttled position, consistent I
with the results of calculation C-1101-211-E610-066and the results of Special Test
Procedure 1-97-0060,"Special Temporary Procedure for Makeup System Changes
Prior to Cycle 12," performed on October 18,1997. !
. in response to the issues discussed in Sections E.8.2 and E.8.3, the licensee
revised the position of the makeup pump suction cross connect valves MU-V-69A
and MU-V-698 from normally closed to open, aligning the suction of all three
makeup pumps to the makeup tank. Previously with the valves normally closed,
only makeup pump 1C was isolated from the makeup tank while pumps 1 A and 1B 1
were aligned to the tank. One of pumps 1 A or 1B functioned for normal reactor
l
coolant system makeup. Safety Evaluation Number 000211-015," Operation with J
MU Suction X-connect Valves Open," was approved on August 25,1997, to
support this change. The safety evaluation concluded that the system configuration
charge was within the bounds of the existing calculation and current licensing basis
and could be made without prior NRC approval. However, the inspectors found that
the safety evaluation failed to consider the effects of a failure of valve MU-V-14A or
MU-V-14B to open. In calculation C-1101-211-E610-066," Makeup Tank Level and
Pressure Limits," Revision 1, developed in support of the safety evaluation, the
licensee did not address the potential single active failure of either BWST supply
valves MU-V-14A or MU V-148 to open under a design basis accident scenario with
offsite power available.' The scenario would result in three makeup pumps operating
with a single flow path from the BWST to the pump suction creating a faster draw
down from makeup tank MUT-1. This condition became more limiting than that
previously analyzed, affecting the basis for the pressure and level operating limits of
the makeup tank.
After questions were raised by the inspectors, the licensee performed further
analysis and determined that this single failure condition was more limiting than the
previously analyzed condition, and that the pressure and level operating limits of the
makeup tank would not bound all design basis conditions. Upon determining this,
the licensee notified the NRC on August 20,1998,in accordance with 10 CFR
50.72(b)(2)(iii) and later issued LER No. 98-09 on September 18,1998. The
licensee stated that if all three high pressure injection pumps were operating in the
high pressure injection mode, and a single failure to open occurred to one of the
two normally closed suction valves, makeup pump suction pressure would be
lowered to the extent that there could be gas entrainment in the pumps' suction in
i
i
!
. .
7
the event of a large break LOCA. The licensee's preliminary review of makeup tank
data determined that, for about one percent of the time during the preceding year,
pressure and level were in the region where gas entrainment could occur. This
potential for gas entrainment in the "C" makeup pump from the makeup tank would
not have existed with the suction cross connect valves closed. Therefore, the
change created the possibility for a malfunction of a different type than any
previously evaluated in the UFSAR.
Later on, a problem related to this issue was identified by the licensee with the
makeup tank level instrumentation. During calibration, the "as found" level
instrument drift was more than expected, causing the indicated level to be higher
than the actual level in the tank. The compounding effect of the instrument drift
was that the potential for gas entrainment to occur increased to about three percent
of the time over the preceding year. The instrumentation issue is discussed in detail
in NRC inspection report 50-289/98-05.
The licensee immediately revised the makeup tank pressure and level operating
limits in Operating Procedure 1104-2, " Makeup and Purification System," Revision
113, to restrict the acceptable range in an operational area that restored the original
design conclusions. The licensee also stated that upon completion of the revised
calculation, additional corrective actions would be taken as required.
The inspectors independently observed that the effect of the error associated with
the ...skeup tank level and pressure limits (Revision 1 of Calculation E610-066) was
to reduce design margin, resulting in less conservatism associated with prediction
for gas entrainment. Calculation E610-66 represented a conservativo analysis of
gas entrainment from the makeup tank which did not specifically account for
(1) different break sizes and RCS back pressure, (2) EOP steps to secure the HPl
pumps, or (3) actual pump configuration and flow rates. The analysis was meant to
bound all design basis cases, thereby minimizing the time at which BWST level is
reduced to where ECCS switchover occurs. The analysis also did not credit
operator actions taken in accordance with EOP 1210-7 (Steps 2.11 and 2.15) to
secure makeup pumps if " throttling criteria" are met.
For the smallest large break LOCA (0.5 square foot break) of concern, the RCS
depressurizes below the DHR pump shutoff head within approximately four minutes.
The EOP throttling criterion of greater than 1000 gpm LPl (DHR) flow per train is
met at this point. Therefore, operators following EOP's willlikely secure at least
one or more makeup pumps in the scenario of interest, and gas entrainment would
be precluded. Further, even if operators failed to follow established EOP's, the HPl
function is no longer needed for ECCS at any time after four minutes in this
scenario. In either case, operators have a 20-25 minute period of time (after this
EOP decision point) to secure or otherwise throttle any one makeup pump before
BWST ievel would reach the point where gas would be predicted at the makeup
tank outlet nozzle. The inspectors also considered that the configuration wherein
one of the BWST MU-V14 valves fails to open would not necessaril; lead to a
__
. .
I
1
1
8
common mode failure of all three pumps due to gas entrainment. This can be seen
by a careful evaluation of the piping isometric and flow paths between the makeup
tank and each pump, as well as the volumetric flow " splits" between the BWST
(95%) and the makeup tank (5%) in a design basis condition. Therefore, the
inspectors judged the safety significance to be limited.
The inspectors assessed the potential for gas entrainment from a risk perspective
and concluded that the change in risk was low. The scenario is only of interest in
certain medium sized LOCAs (and in steam generator tube rupture) where offsite
power remains available, the makeup tank is outside of its operational pressure / level
limits (which for the case in question represented 1-3% of the period
October 1997 - August 1998), a BWST supply valve fails to open, and certain
proceduralized operator actions (e.g., secure HPI, isolate tank) are not credited.
However, notwithstanding the risk and technical significance of tha change made, i
the approval of SE 211-15 on August 25,1997, to cross-tie the HPl header
introduced an unreviewed safety question, when the potential to fail the "C"
makeup pump due to gas entrainment was created. Although the makeup tank
preserve / level operational limits were revised, the unreviewed safety question still
existed representing a violation of 10 CFR 50.59 (a)(2)( I). (VIO 98-06-01)
c. Conclusions
The corrective actions associated with the adequacy of makeup tank pressure / level
curves adequately addressed the original AE team findings. However, a violation of
10 CFR 50.59 was identified with the opening of the HPl suction header cross
connect valves without NRC approval. The safety evaluation failed to identify that
the change involved an unreviewed safety question, as the change introduced a
malfunction of a different type than any evaluated in the UFSAR;i.e., a failure of
the "C" HPl pump due to gas entrainment.
E8.5 (CLOSED) URI 50-289/96-201-05.Desian Basis Valve Stroke Times in Surveillance
Procedure
a. Inspection Scope
The AE team identified a concern involving the stroke times associated with Makeup
pump recirculation isolation valves MU-V-36&37, and injection isolation valves MU-
V-16A through D. The referenced design basis stroke times for the valves in
surveillance procedure 1300-3H,"lST of MU Pumps and Valves," of 36 seconds
was higher than the actual design basis ECCS delay time of 35 seconds assumed in
the LOCA analysis.
l
!
,
. - - . - - . - - - . - _ - - .. - - _ - - - - . - - - - - - -
- *
- j
!
i
9 i
b. Observations and Findinas
FSAR section 14.2-14 (Accident Analysis) states that the ECCS delay time (LOCA) I
was 35 seconds based on 1 second Calculation E610-66 represented a
conservative analysis of gas entrainment from the for instrumentation lag,10 l
'
seconds for start of the emergency power source if offsite power is not available, j
and 24 seconds for system response time. However, FSAR 6.1.3.1, (ECCS) states
!
that systems respond within 25 seconds, and procedure 13OO-3H had 25 seconds .
j
as criterion for the valves' stroke times consistent with FSAR 6.1.3.1. :
!
"
In a recent analysis to support a request for power up rate, the licensee had ,
concluded that 25 seconds was an acceptable system response time. The NRC !
issued amendment No. 203 which approved the LOCA analysis for the impending
power up rate to 2620 megawatts thermal, TMI-12772 MWT ECCS Analysis with !
, RELAP5/ MOD 2-B&W, (Doc. No. 43-10222P), Revision 0, dated March 1997.
Section 4.2.1.9 (Vol. I, Large Break, Vol. II, Small Break), ECCS, of the analysis
assumes a delay time of 36 seconds for HPI (and LPI) injection based on: signal
generation, electrical supply startup, and initiation of the pumped injection flows. )
Since the approved design bases was then 36 seconds, the inspector determined !
that the original discrepancy was resolved and no longer a concern
t
The inspector reviewed surveillance procedure 13OO-3H,"lST of MU Pumps and l
Valves," revision 49, and verified that the stroke times specified for the valves were i
less than 25 seconds (18 open and 19 closed) and the design basis stroke time I
<
referenced was 25 seconds. The inspector reviewed past test data and verified that i
'
the valves all stroked within 15 seconds eliminating any past operability concern.
'
c. Conclusions
The licensee demonstrated that the stroke times for makeup system valves
specified in surveillance procedure 13OO-3H,"lST of Makeup Pumps and Valves,"
- were appropriate. t
'
E8.6- (CLOSED) IFl 50-289/96-201-06,Consecuences of Failure of Auxiliarv Steam Pioina
.
[ a. Insoection Scooe
- -
+
The AE team identified a discrepancy involving the non safety-related auxiliary
steam piping in the auxiliary building. The licensee had no analysis to show that a
'
failure of the piping due to a seismic event would not result in degradation of
safety-related equipment classified for mild environment in the area.
i
L
)
i
.
] '
- - n, - - - - _ - - . ,,- - - , . -
.-_ .- . - .. --
< .
i
10 !
l
b. Observations and Findinas i
To resolve this item, the license performed a seismic analysis of the auxiliary steam
pipes. The inspector reviewed report 240046-R-001," Seismic Verification Review
of Auxiliary Steam System Piping Three Mile Island Unit 1," Revision 1, dated
June 5,1998. The report, prepared for GPU Nuclear by EQE Engineering
Consultants, concluded that based on walkdowns and confirmatory analysis, the ,
auxiliary steam piping has sufficient seismic margin to maintain pressure integrity l
during and following a seismic event equal to the plant safe shutdown earthquake
(SSE). The seismic verification was based on a methodology analogous to that
used for seismic verification of mechanical equipment under the USl A-46 program.
Since no physical changes were required, the isst.e had not represented any '
operational concern. The inspectors also concluded that the methodology used
provided a reasonable framework to assess the effects of a seismic event on the
nonsafety-related auxiliary steam piping.
c. Conclusions
The licensee adequately demonstrated that during a seismic event, the auxiliary
steam piping will not fail and since no physical changes were required to arrive at
this conclusion, the issue did not represent any past operational concern.
E8.7 (CLOSED) IFl 50-289/96-201-07, Loss of Pressure in MU&P Tank due to Letdown
Line Break
a. Inspection Scope
The AE team identified a discrepancy involving the available net positive suction
head (NPSH) for the makeup pump. The impact of loss of pressure in makeup tank
MU-T-1 on the available NPSH for the makeup pumps due to a postulated letdown
line break or crack combined with the failure of the check valve in the line to the
makeup tank was not clearly addressed. l
b. Observations and Findinas
The licensee performed an evaluation of the available NPSH for the makeup pumps
if the gas pressure in the makeup tank were lost due to a pipe break This
evaluation concluded that adequate NPSH would be available to operate one
makeup pump, based on zero psig tank pressure and normal tank level.
The inspector reviewed the analysis that supported the licensee's conclusion and ;
found it appropriate. The licensee stated that only one makeup pump would be
required to operate during this event since the pipe break would have rendered one
safety train inoperable. The inspector reviewed the possible scenarios and
concurred with the licensee's determination. The inspector also reviewed the
licensee's evaluation and verified that adequate NPSH would be available for a
makeup pump with no gas pressure in the makeup tank.
1
. - . . . .-. -. .- - - - - . - . . - . - - - . . . . - - _ . - - - . . . - . - - .
. , i
t
r
i
11 -
i
c. Conclusions J
. -.-The licensee adequately demonstrated that with the makeup tank depressurized, or f
with a low-low level in the BWST, there was no safety concern with the makeup !
pumps' NPSH.
E8.8 (CLOSED) IFl 50-289/96-201-08.M&UP Pumo NPSH When Takina Suction From {
BWST or Makeun Tank l
l
'
a. Insoection Scope
l
. The AE team identified a discrepancy with the available net positive suction head I
(NPSH) for the makeup pumps when taking suction from the BWST at low-low level
or during potential vacuum conditions in the makeup tank. !
!
b. Qbservations and Findinas !
I
The licensee completed calculation C-1101-211-E610-066," Makeup Tank Level
and Pressure Limits," Revision 1, to address the available NPSH for the makeup i
pumps under accident conditions. This calculation established the required pressure !
and level operating limits for the makeup tank, and considered the minimum !
indicated BWST level of 9'-6" prior to transferring the makeup pumps to " piggy- j
back" mode. The licensee stated that with the makeup tank maintained in the i
normal operating band there was no specific mechanism or sequence of ovents that l
could result in a vacuum in the makeup tank during draw down. In addition, i
operator was required to open the suction valves from the BWST if the makeup tank l
reached 55 inches. The licensee concluded that the potential for vacuum conditions ;
in the makeup tank was precluded by system design and operational controls. j
!
The inspector reviewed calculetion C-1101-211-E610-066," Makeup Tank Level and !
Pressure Limits," and verified . hat adequate NPSH would be available for the .
makeup pumps based on the required pressure and level operating limits for the i
makeup tank being maintained. The inspector reviewed Abnormal Transient l
Procedure (ATP) 1210-6, "Small Break LOCA Cooldown," Revision 25; ATP 1210-
7, "Large Break LOCA Cooldown," Revision 25; and 1210-10," Abnormal !
Transients Rules, Guides and Graphs," Revision 34 and verified that " piggy-back" !
mode would be established prior to the indicated BWST level reaching 9'-6". The l
inspector also reviewed ATP 1210-1," Reactor Trip," Revision 37 and Alarm j
Response Procedure D-3-2, " Main Annunciator Panel D," Revision 13 and verified
that the suction valves from the BWST, MU V-14A and MU-V-148, would be i
opened if the makeup tank reached 55 inches. j
.c. Conclusions -
l
'
l
'
The licensee adequately demonstrated that with the makeup tank depressurized, or
with a low-low level in the BWST, there was no safety concern with the makeup ,
pumps' NPSH when taking suction from the BWST. )
> >
'
l
1
1
l
1
.-,- , . . . , . _ . ,.
___ _ _ _ . - . - . _ _ . _ _ _ . _ . _ -._ _ . _ . _..__ . ._ , _ . _ . _ _ _ . _
- .
12 ,
i
,
E8.9 (CLOSED) IFl 50-289/96-201-09,Incomolete DC System Voltaae Droo Calculations I
!
a. Insoection Scoos -
l
i
The AE team identified a discrepancy with the voltage drop analysis for makeup l
valve MU-V-18. Specifically, the voltage drop analysis was not performed due to !
lack of design data for the valve and for other components listed in the existing 125 (
Vdc sys.em calculation, C-111-734 5350-OO4. !
b. Qhteryations and Findinas
To address this issue, the licensee re-evaluated the minimum available voltage for
all the components in revision 2 of 125 Vdc system calculation, C-111-734-5350-
004. The inspector noted that the licensee had obtained the required input data of ,
sona components, from manufacturers (specification data sheets) and in some i
cases utilized plant test data. The cable impedance data was used from design
documents. The inspector reviewed the. voltage drop analysis and found no :
problems. For exampfe, a calculated minimum of 102 volts was available at the
' ,
'
terminal of the makeup isolation valve MU-V-18 compared to a minimum required i
voltage of 90 volts specified by the vendor, under the worst case loading condition, I
and assuming a worst case voltage of 105 V at the battery. The inspector verified !
that other components such as solenoid operated relief valve RC-RV2 and main
. steam motor operated valves, MS-V-10A and B, had sufficient voltage to perform i
their intended design functions. j
c. Conclusions
l
t
Voltage drop analysis and calculations showed that safety related equipment, i
including valve MU-V-18 would have sufficient voltage to perform their design i
function.
E8.10 (CLOSED) URI 50-289/96-201-10 Alianment of MU-V-18 DC Power Suoolv i
a. Insoection Scope
The A/E team identified that, the licensee had not established any instruction or ;
procedures to assure that the makeup system discharge cross-connect valves (MU- ;
V-76-A,B and MU-V-77A and B) and the dc power source of makeup system valve l
MU V-18 were properly aligned. i
>
'
b. Observations and Findinas
The team identified that on January 8,1998, the de power source of makeup
system valve MU-V-18 was not properly aligned. With the valve mechanically
'
aligned as a loop "A" valve, it was electrically powered from the "B" 125 VDC
battery source. Under normal:lignment of HPl system, MU-V-18 supply is powered
from "A" de distribution panel (1M). The licensee corrected the DC power
, alignment immediately, and issued a temporary order change that required periodic ;
g--,p -t---*- -'*t - - - - - v-+w- wt - - --- - * - " ' '
. .
13
verification of system alignment by control roorn staff to assure its proper operation. ;
Proper mechanical /clectrical alignment of the valve is required because the loss of a
125 VDC power train, which would render a HPl train inoperable, could result in
failure of MU-V-18 to close, thus, degrading the other HPl train. Therefore, the
selection of the "A" or the "B" 125 VDC power source from the 1M,125 Vdc
distribution panel must be based on the makeup pump discht,rge lineup.
At that time of the AE team inspection, FSAR section 6.1.3.1, stated that in the
event of a small break LOCA (SBLOCA) a 70/30 HPl flow split (70% to the reactor
core,30% out of the break) will assure adequate core cooling. This 70/30 flow
split is ensured by closing air operated valve MU-V-18, which is a safety-related
valve that automatically isolates on an engineered safeguards (ES) actuation signal,
thus isolating the makeup line flow to HPl leg "B" downstream of the cavitating
venturi. Valve MU-V-18 is powered from 125 VDC distribution panel 1M, which in
turn is supplied from either of the redundant 125 VDC main distribution panels 1 A
or 18 via an automatic transfer switch. The transfer switch is designed to be
locked out when an ES signalis received. The design of the control scheme for
valve MU-V-18, is such that a loss of 125 VDC power to the valve would not close
the valve if control air was not available. The valve can be remote closed if both
power and control air were available. FSAR analysis also took credit for partial
isolation of the makeup flow if the operator closes the non safety-related makeup
control valve MU-V-17 (upstream of MU-V-18). During the A/E team inspection, the
licensee explained that even though MU-V-17 was not a safety-related component it
was included in the inservice test program, and its operational status was monitored
continuously because the valve was used for controlling the makeup flow during
normal operation.
The licensee further determined thm, assuming an ES actuation concurrent with,
prior to, and after the de power lou, there was no adverse impact on nuclear
safety. In the first and second cast s, both trains of HPi will be initiated and then
MU-V-18 fails "open" when dc is lor.t. The running HPI pump does not trip when
dc is lost. With two HPl pumps operating the flow through MU-V-18 does not
prevent HPl from performing its nuclear safety function. In the third case, the
normal de distribution panel power supply (IM) would be transferred to the alternate
de source when de is lost and MU-V-18 will close upon ES actuation, although only
one HPI pump will start in this case. Since one HPl pump is capable of providing
the minimum HPl flow assumed in the ECCS LOCA analysis, therefore, there was no
adverse impact on nuclear safety. The inspector reviewed this analysis and
identified no concern.
Based on the above review, the inspector concluded, that prior to January 8,1997,
GPUN did not establish adequate instructions or procedure for the proper alignment
of the de power source of makeup system valve MU-V-18. As a result, the
alignment of the HPl system and the de power supply of distribution panel (1M) was
not satisfactorily accomplished on January 8,1997. This was contrary to 10 CFR
50, Appendix B, Criterion V, " Instructions, Procedures, and Drawings." However,
this failure constitutes a violation of minor significance and is not subject to formal
enforcement action.
.
. - - - . - - - . - _ - . - - - - - . . . - - . - - . - . - . - - . . - -
- e
l
!
l
14
I c. Conclusions
-
Although operation of makeup system valve MU-V-18 prior to January 8,1997,
l was not appropriately controlled by procedures, it did not represent a more broad
concern with respect to establishment of prococures for Esfety activities. This
minor violation was not subject to formal enforcomen'. action,
i
! ' E8.11 (CLOSED) URI 50-289/96-201-11.Makeuo Tank Levet instrument Looo Tolerances
'
a. Insoection Scope
l
L The A/E team identified, several instances in which the licensee had used non-
conservative instrument loop errors in data sheets of calibration surveillance
procedures. The calculated allowable loop errors were more restrictive than those
in the instrument loop data sheets.
'
b. Observations and Findinas
l To address this concern, the licensee revised applicable plant calculations and
! surveillance procedure data sheets. The inspector reviewed the updated instrument
data sheets and found that the loop errors were equal to or more than the estimates
in the revised calculations. For example, a specified loop error of *1% shown in
data sheet of level recorder MU-14-LR for make-up tank level instrumentation (MU-
14-LT)in surveillance procedure was found conservative, compared to 12.74%
estimated in revised loop error calculation (C-1101-662-5350-049, Revision 5).
Similarly, the review of other data sheet for the control room indication instrument
j loop indicator, MU-Ll-778A, and computer point A0498 associated with instrument
loop, MU14-LT, were found conservative and consistent with the revised
calculation. The inspector concluded that the licensee's use of a conservative loop
error considered in data sheets of established surveillance procedures in above
cases was appropriate. The inspector also verified that none of the discrepancies
previously identified had any effect on equipment functionality. !
c. Conclusions
inconsistencies identified in data sheets of calibration surveillance procedures were
resolved. The instrument loop errors in data sheets of calibration procedures
reviewed were found to be acceptable.
l
E8.12 (CLOSED) URI 50-289/96-201-12.FSAR Discreoancies
!
a. Insoection Scope
- The AE team identified discrepancies in various sections of the FSAR regarding the
l makeup system. While the discrepancies were minor in nature, the team was
,
concerned with the fact that they had not been corrected and the FSAR updated to
'
assura that the information included in the FSAR contained the latest material as
required by 10 CFR 50.71(e).
. _. _- , -
. - -
- - . . - -- - - _ .
_ . . _ .. _ _ _ _ _ _ _ _ _ _ _ _ ._._ _ _ _ _
. ,
15
b. Observations and Findinas
in subsequent follow up, the NRC determined that, although a violation of 10 CFR
50.71(e), these discrepancies were of minor significance and the issue was
therefore identified as a non-cited violation in a letter dated October 8,1997 to the
Licensee.
During this inspection, the inspector reviewed the licensee's actions taken to
correct the identified discrepancies as well as those taken to prevent similar FSAR
discrepancies. All the discrepancies identified in the AE team report were corrected
in the April 1998 FSAR update. In a broader sense, the licensee established FSAR
section owners to improve the quality of the FSAR and also transferred the
ownership of the FSAR to the engineering department. The inspector did not
identify any discrepancies in the portions of the FSAR reviewed during this
inspection.
c. Conclusions
The licensee adequately resolved the FSAR discrepancies identified by the AE team.
In addition, the ongoing efforts by the licensee to improve the overall quality of the
FSAR appeared appropriate.
E8.13 (CLOSED) URI 50-289/96-201-13.and eel 97-070-01013. Adeausev of BWST
Setooint for DHRS Pumo Switch over to RB Sumo
a. Inspection Scope j
l
'
The AE team identified concerns associated with calculation C-1101-212-5310-
050, "TMI-1 BWST Vortexing Determination," Revision O. This calculation l
evaluated the DHRS pump suction transition from the BWST to the reactor building
(RB) sump under accident conditions. Following further NRC review, this issue was ,
identified as a violation of 10 CFR 50, Appendix B, Criterion Ill, " Design Contrc'," {
as documented in a Notice of Violation and imposition of Civil Penalties, dated !
October 8,1997.
b. Observations and Findinas
The concerns included (1) failure to consider instrument uncertainty in the low-low
level alarm setpoint; (2) the use of a non-conservative reactor building pressure
assumption; (3) the validity of the assumed time for operator response; (4)
disagreement between the valve stroke times used in the calculation and the design
basis times stated in the valve test procedure; and (5) incorrect methodology used
for vortex determination. Based on these concerns the team concluded that there
was a potential for air entrainment in the DHRS that could have rendered the pumps
... - . . - . - - - . - - - . - - - . - . - . - - - - - - ~ . _
- b
I
!
16
l
The licensee performed an operability evaluation of the DHRS on
December 20-21,1996, concluded that both trains of the DHRS and the RB spray ;
. system were inoperable, and notified the NRC in accordance with 10 CFR 50.72. l
The licensee issued temporary changes to several procedures to require the ;
operators to begin transferring the DHRS pump suction from the BWST to the
reactor sump when the BWST reaches the low level alarm setpoint of 9'-6", and to l
complete the transfer when the BWST reaches the low-low level alarm setpoint of
l
6'-4". Based on the changes the DHRS and the RB spray system were considered j
The licensee issued calculation C-1101-212-5310-050,"TMl-1 BWST Vortexing :
Determination," to address the errors identified during the design inspection. In i
addition, the licensee revised Abnormal Transient Procedure (ATP) 1210-6, "Small l
Break LOCA Cooldown," Revision 25; ATP 1210-7,"Large Break LOCA Cooldown,"
Revision 25; ATP 1210-8,"RCS Superheated," Revision 22; and 1210-10, l
'
, " Abnormal Transients Rules, Guides and Graphs," Revision 34 to require the
operators to begin transferring the DHRS pump suction from the BWST to the
reactor sump when the BWST reaches the low level alarm setpoint of 9'-6", and to I
complete the transfer when the BWST reaches the low-low level alarm setpoint of )
g 6 '-4 ".
The inspector reviewed Calculation C-1101-212-5310-050,"TMI-1 BWST Vortexing -
l Determination,"_ Revision 1 and verified that the errors identified during the design I
inspection had been corrected, and that the calculation inputs, assumptions, and
J
, method were consistent with the design bases. ,
'
The inspector reviewed Abnormal Transient Procedure (ATP) 1210-6, "Small Break
, LOCA Cooldown," Revision 25; ATP 1210-7,"Large Break LOCA Cooldown,"
Revision 25; ATP 1210-8,"RCS Superheated," Revision 22; and 1210-10,
, " Abnormal Transients Rules, Guides and Graphs," Revision 34 to verify that these
I procedures were consistent with the calculation. The inspector also reviewed
i
Surveillance Procedure 1300-3B,"lST of DH-P1 A/B and Valves," Revision 60 and
.
Surveillance Procedure 1303-11.16,"DH Leak Test," Revision 35 and verified that
?~
the critical valve stroke times used in the calculation were consistent with the
'
design basis stroke times.
2 c. Conclusions
I The issues associated with potential air entrainment into the DHRS pump suction
during switch over from the BWST to the RB sump have been appropriately
- resolved. The corrective actions taken to address the violations stemming from this
<
issue were adequate.
]
j
t- )
i
k
'
l
-
l
)
-
-w,y =e-- --
.ye w e-- e , e - a w-n
L _
,
,
i
l
17
E8.14 (CLOSED) URI 50-289/96-201-14.and eel 97-070-01093. Adecuacy of Safety
Evaluation for an FSAR Chanae Creditina Containment Overoressure for LPl Pumos
a. Insoection Scope
The AE team identified concerns associated with calculation C-1101-212-5360-
027, "NPSH Available for LPl and BS Pumps Following Large Break LOCA," Revision
- O and Safety Evaluation SE-115403-004, Revision O. Following further NRC
review, this issue was identified as a violation of 10 CFR 50.59, Changes, Test and
Experiments, as documented in a Notice of Violation and imposition of Civil
Penalties, dated October 8,1997.
b. Observations and Findinas
Calculation C-1101-212-5360-027 determined that the NPSH requirements for the
DHRS pumps would not have been met for a LPI flow of 3300 gpm as indicated by
flow instrumentation under post-LOCA conditions unless credit was taken for sub-
cooling of the reactor building sump water (i.e., containment overpressure).
l
The NRC safety evaluation report (SER) for TMI-1, dated July 11,1973, stated that
the licensee's NPSH evaluation was based on the assumption that reactor building
pressure was equal to the vapor pressure of the sump water and was acceptable.
In this evaluation containment overpressure was not considered.
A statement was added to FSAR Section 6.4.2, Update 9, dated July 1990, to
address sump sub-cooling. It stated, "an additional calculation indicated that
- sufficient NPSH would be available for the maximum flows, as limited by procedure,
'
including instrument errors, recirculation flow and taking credit for sump sub-cooling
(containment overpressure)." The licensee's safety evaluation, SE-115403-004,
Revision 0, failed to identify that, because the required NPSH for these pumps could
not be met without taking credit for containment overpressure, a potential
unreviewed safety question was involved.
l
The licensee issued calculation C-1101-210-E610-011,"LPI and BS Pump NPSH
Available from the RB Sump Following a LBLOCA," Revision 1 to determine the
maximum allowable indicated LPI and BS pump flows, considering instrument
uncertainty, to ensure that adequate pump NPSH is available, with no credit taken
for containment overpressure in addition, the licensee revised Abnormal Transient
Procedure (ATP) 1210-6,"Small Break LOCA Cooldown," Revision 25 and ATP
1210-7,"Large Break LOCA Cooldown," Revision 25 to limit the LPI and BS flow
rates.
l
The inspector reviewed calculation C-1101-210-E610-011,"LPI and BS Pump NPSH
Available from the RB Sump Following a LBLOCA," Revision 1 and verified that no
credit had been taken for containment overpressure, and that the calculation inputs,
assumptions, and method were consistent with the design bases. The inspector
also reviewed related calculations C-1101-210-E610-010," Reactor Building
l
l
l
!
. _ _ ~ -. ~ . - - . - . . . . - . - _ , . . . - _ - . - . - . - . . .. - .- - -. - . -
. ,
18
Minimum Level During Recirculation Following a LBLOCA," Revision-1 and C-1101-
212-5450-040,"BWST Minimum Usable Volume," Revision 2 to verify that these j
calculations were consistent with calculation C-1101-210-E610-011. The inspector i
reviewed Abnormal Transient Procedure'(ATP) 1210-6,"Small Break LOCA
Cooldown," Revision 25 and ATP 1210-7,"Large Break LOCA Cooldown," Revision
25 to verify that these procedures included flow limits that were consistent with the
calculation C-1101-210-E610-011.
c. Conclusions j
The issues associated with improperly taking credit for containment overpressure in ;
LPI pumps NPSH calculations were resolved. The corrective actions taken to i
address the violation stemming from this issue were appropriate. Calculations
properly reflected that adequate NPSH would be available for the LPl pumps without ,
taking credit for containment overpressure. I
i
E8.15 (CLOSED) IFl 50-289/96-201-15. Technical Soecification Discreoancy
a. Inspection Scone !
The AE team identified a discrepancy in the technical specification involving the -;
DHRS. Specifically, TS 3.3.1.1.f stated that the two reactor building sump isolation ;
valves (DH-V-6A&B) shall be manually or remotely operable. However, only remote ;
operation of the valves would support the design basis assumptions for the l
transition time of the DH pump suction switch over from the BWST to the RB sump )
during post-accident conditions. l
.
\
i
b. Observations and Findinas i
4
1
The licensee initiated technical specification change request (TSCR) No. 263 to
delete the " manual" operation from the technical specifications (TS). In TS 1
Amendment No. 203, the " manual" was deleted. The valve surveillance tests
required that the valves be tested for remote operation and the licensee stated that
TMl-1 had not been operated with these valves in a condition where they were only
. manually operable. The inspector verified these changes and found the licenst e's
,
actions appropriate.
l
!' c. Conclusions I
The inspector found this discrepancy to be of minor significance and determimd
.that the licensee had taken appropriate actions to correct it. This item is closed.
l
l
,
,
4
1
$
,, -. . . - , -- .
-
. - - . . - . - - . - ~ . - - . - . - . - - . _ . . - - . - - . - . ~ ~ . - _
- '
! l
i
i
1
19 .
1
E8.161 CLOSED) URI 50-289/96-201-16. Discrepancy Between FSAR and Technical
Specifications Reaardina DHRS Leakane I
a. Insoection Scope
i
- The AE team identified a discrepancy between Technical Specification 4.5.4, DHRS
l leakage, and FSAR Section 6.4.4 and Table 6.4-3. The discrepancy was that the !
l technical specification allowed 6 gal /hr leakage from the DHRS while the FSAR ;
l specified the leakage for the DHRS and Building Spray system as 2255 cm /hr (0.6 - i
gal /hr).
i
b. Observations and Findinas
l To resolve the discrepancy, the license revised the technical specification j
!
(Amendment No. 205) and applicable plant procedures to reflect the design bases {
leakage number (0.6 gal /hr) specified in the FSAR. In an initial review, the licensee '
determined that the discrepancy was not safety significant because the dose of
0.30 rem (corresponding to the TS 6 gal /hr leakage) still represented a negligible ;
. portion of the total two-hour thyroid dose limit of 189 rem.
The inspector reviewed the changes in the technical specification and in the
following procedures: SP 1303-11.16,"DH Leak Test," Revision 35; SP 1303- ;
11.27, "MU Pump Suction Leak Check," Revision 9; SP 1303-11.32,"MU Pump
Discharge and Letdown Leak Check," Revision 0; and SP 1303-11.50,"RB Spray i
Leak Check," Revision 6. The inspector found a discrepancy in procedure SP 1303- '
11.16, which still reflected an acceptance criterion of less than or equal to L<_) 6
gal /hr for DHRS leakage. The licensee indicated that since the procedure would
l
'
only be used during outages, the appropriate change was not incorporated yet. The 1
. inspector found this inappropriate since an approved plant procedure that was i
obviously in conflict with the technical specification was being maintained. The
procedure was then revised by the licensee such that no specific acceptance
criterion was specified but instead reference was made to verify that the leakage I
was in compliance with the tech specs. With this change, the procedure was in
compliance with the technical specification, and further changes to the technical
specification would not require any changes to the pacedure. The inspector
reviewed the revised procedure and identified no other concerns. '
l. c. Conclusions
l l
The licensee was initially untimely at correcting the discrepancy in procedure SP- ,
1303-11.16,"DH Leak Test." Nevertheless, the discrepancy between the FSAR and '
- the technical specification was adequately resolved.
i
4
J
.
t
I
.-
4
-, . -, .,- - - - _ . - - - , - _, ,
__ _ _.
1 . .
20
E8.17a (CLOSED) URI 50-289/96-201-17A.and eel 97-070-10143, Leak Testina of
DHRS Pumo Suction Check Valves
a. Insoection Scoce
The AE team identified a concern with the licensee's IST Program, dated March
1,1994. The leak testing of decay heat (DH) removal system (DHRS) check valves
DH-V-14A and DH-V-148 was not included in the program. Following further NRC
review, this issue was identified as a violation of 10 CFR 50, Appendix B, Criterion
Ill, " Design Control," as documented in a Notice of Violation and imposition of Civil
Penalties, dated October 8,1997.
b. Observations and Findinas
These valves are in the suction lines from the Borated Water Storage Tank (BWST)
to the DHRS pumps. Leakage of post-accident reactor building sump water through
these valves in conjunction with leakage through the isolation valve or the failure of
the isolation valve to close could have resulted in increased offsite dose.
In a licensee memorandum (5310-92-024,"DH-V-14A/B, DH-V-5A/B, BS-V-52A/B
Valves), a calculation incorrectly concluded that check valves DH-V-14A and DH-V-
148 did not perform a safety function in the closed position, because an incorrect
value for the reactor budding pressure was used. Consequently, leak testing of
check valves DH-V-14A and CH-V-14B was not included in the IST program. The
licensee attributed the cause of the violation to lack of familiarity and inadequate
training on the requirements of procedures, as well as lack of management
enforcement to fully implement procedural requirements. The licensee established
interim operating restrictions on reactor building pressure, BWST temperature, and
reactor building fan cooler availability on December 20,1996, to assure that
leakage through these check valves could not reach the environment. The licensee
issued Surveillance Procedure 1300-3Z.5,"lST of ECCS Bypass Valves - DH
Suction Valves," Revision O to test the leakage through check valves DH-V-14A
and DH-V-148.
The inspector reviewed Surveillance Procedure 1300-3Z.5,"lST of ECCS Bypass
Valves - DH Suction Valves," Revision O to verify that check valves DH-V-14A and
DH-V-14B were being tested in accordance with the requirements of 10 CFR
50.55a. The inspector selected some other safety related valves during the course
of this inspection and verified that the valves were included in the IST program and
tested appropriately for their safety functions.
c. Conclusions
Appropriate actions were taken to address IST program discrepancies involving
DHRS valves. The valves in question were properly included in the IST program.
i
+ ,
l
21
E8.17b (CLOSED) URI 50-289/96-201-178.and eel 97-070-08014. Leak Testina of
DHRS Pumo Discharae Check Valves
a. Insoection Scope
l
l
The AE team identified a concern associated with testing of the DHRS pump
discharge check valves, DH-V-16A and DH-V-16B. The IST program did not include
the closed function of DH-V-16A and DH-V-16B. Following further NRC review,
this issue was identified as a violation of 10 CFR 50, Appendix B, Criterion XI,
" Test Control," and documented in a Notice of Violation and Imposition of Civil
Penalties, dated October 8,1997.
b. Observations and Findinas
if one of the DHRS pumps failed after an accident, discharge cross connect valves
DH-V-38A and DH-V-388 would be opened to provide flow through both low
pressure injection paths. In this condition the discharge check valve associated
with the failed pump, DH-V-16A or DH-V-168, would have to close to ensure
adequate low pressure injection flow.
The licensee determined that the check valves (DH-V-16A and B) were not included
in the IST program because the determination of the safety functions for the valves
was inadequate. The licensee performed a test on December 19,1996, during the
design inspection, to assure that check valves DH-V-16A and DH-V-168 seated in
the closed position. Both valves performed acceptably during the test. The
licensee revised Surveillance Procedure 1300-3P,"lST of Check Valves During
Shutdown," Revision 20 to include testing of check valves DH-V-16A and DH-V-
16B in the closed position. The licensee also performed a review of IST program
scope and implementation. The results of the review did not reveal any
programmatic concern with the IST program.
The inspector reviewed Surveillance Procedure 1300-3P,"lST of Check Valves
During Shutdown," Revision 20 to verify that check valves DH-V-16A and DH-V-
16B were being tested in the closed position. In addition the inspector reviewed
Special Temporary Procedure 1-96-0060," Closure Test of DH-V-16A/B," dated
December 19,1996 to verify that DH-V-16A and DH-V-16B seated in the closed
position on December 19,1996. The inspector also reviewed and discussed the
results of the IST program reviews conducted by the licensee. No programmatic
issues were identified.
c. Conclusions
l Appropriate actions were taken to address IST program discrepancies involving
DHRS valves. The valves in question were properly included in the IST program.
_ . _ _ ._
. ,
22
E8.17c (CLOSED) URI 50-289/96-201-17C,and eel 97-070-08014,Insoection of
DHRS Pumo Vault Floor Drain Check Valves
a. Inspectionycope
The AE team identified concerns associated with testing of the DH vault floor drain
check valves. The IST program and the inspection procedures for check valves did
not include the ball float type check valves installed in the DH vault floor drains.
Following further NRC review, this issue was identified as a violation of 10 CFR 50,
Appendix B, Criterion XI, " Test Control," as documented in a Notice of Violation
and Imposition of Civil Penalties, dated October 8,1997.
b. Observations and Findinas
FSAR section 6.4.5 stated that a leak in a DH vault would not affect more than one
train of equipment because there are check valves in the floor drains to prevent
water from backing up into the vaults. There were two floor drains from each DH
vault to the auxiliary building sump.
The licensee inspected these check valves during the AE team inspectior, and
determined that the valves were operable. The licensee revised preventive
maintenance procedure U-17, "Zurn Floor Drain inspection," Revision 5 to include 1
inspection of ball float type check valves. In addition the licensee stated that these
check valves would be added to the IST program as augmented IST, and that i
'
Drawing 302-719," Sump Pump & Drainage System (Reactor & Auxiliary Building)
Flow Diagram" would be revised to identify these drains.
The inspector reviewed preventive maintenance procedure U-17, "Zurn Floor Drain
l
Inspection," Revision 5: Surveillance Procedure 1300-4G,"BS & DH Floor Drain ;
inspection," Revision 0; and administrative procedure 1041, "lST Program j
Requirements,' Revision 31 to verify that the DH vault floor drain check valves
were included in the IST program as augmented IST. In addition the inspector 1
reviewed Drawing 302-719," Sump Pump & Drainage System (Reactor & Auxiliary
Building) Flow Diagram," Revision 56 to verify that the DH vault drains were
identified.
c. Conclusions
Appropriate actions were taken to address IST program discrepancies involving
DHRS valves. The valves in question were properly included in the IST program, i
,
. . .
.-
! i
, 8
,
. 1
!
23 ,
1
E8.18 (CLOSED) URI 50-289/96-201-18,and eel 97-070-03013 Timeliness of Action on
- SSFl Open items
l
l ;
a .~ lnspection Scone !
i
!
. The AE team identified 15 items from a safety system functionalinspection '
. performed by the licensee in 1992 that were still open. Following further NRC l
review, the issue was identified as a violation of 10 CFR 50, Appendix B, Criterion l
XVI, Corrective Actions, and documented in a Notice of Violation and Imposition of )
Civil Penalties, dated October 8,1997.
b. Observations and Findinas ;
1
!
. The licensee determined that the main reason for the violation was management's j
tolerance of allowing longstanding technicalissues to remain open. When the SSFl l
was conducted in 1992, procedural controls had not been established to ensure
i- that such adverse conditions would be resolved in a timely manner. The licensee
! entered the SSFl and other open items into their corrective action program (CAP).
p
The inspector reviewed the licensee's tracking system and verified that all 15 SSFl )
open items (212-1 through 13, and 212-42 and 61) identified by the AE team had i
been appropriately resolved. The inspector reviewed some of the resolutions and 1
identified no additional concems. In addition, the inspector reviewed the licensee's )
actions taken to address some open items from a Service Water System Operation :
L Performance inspection (SWSOPI) performed in 1995. All but five of the thirty :
l three SWSOPl open items reviewed were closed. The inspector reviewed the status ;
of the five remaining items and determined that they were being addressed
appropriately and did not present any safety concern. For example, a TDR (1190) l
l to address potential single failure vulnerability with the Decay River / Decay Closed
Cooling Water had been completed and verified by an independent group
(Contractor), but had not been verified and signed off yet by management due to ,
existing workloads and other priorities. The results showed that no single failure i
vulnerability existed. l
l !
c. Conclusions
i
L - Open items associated with SSFl and SWSOPl activities were appropriately entered ,
into the corrective action program and being adequately resolved. :
l E8.19 (CLOSED) IFl 50-289/96-201-19, Evaluation of Reactor Buildina Sumo Screen for
Reverse Flow
I
a. Insoection Scope l
t
.
!
The AE team identified a concern with the capability of the reactor building sump
i screens to withstand pressures due to reverse flow if the RCS drop line was opened l
during post-accident conditions to control boron precipitation in the reactor core.
i
,
,
'
. .. .. ,
. . __ . - _- - . . -
,
. ,
l
'
l
24
b. Observations and Findinas
l
At the time of the AE team inspection, the licensee stated that an analysis to ;
determine the acceptable RCS pressure below which the drop line could be opened I
without damaging the reactor building sumps had not been performed and that an
evaluation would be performed. The licensee performed a preliminary operability
determination of this issue, documented in Plant Review Group (PRG) Meeting
Number 97-046, held on June 26,1997. They concluded that there was a
reasonable expectation that the potentially affected systems would remain operable.
After further investigation the licensee revised the system operation to eliminate the
use of a reverse flow path through the reactor building sump screens during post-
accident conditions. This change was documented by CAP Number T1997-0439,
dated June 27,1997.
i
The inspector found the licensee's actions taken to address this issue appropriate. l
The inspector reviewed the initial operability determination and found that it
provided reasonable basis for concluding that the affected systems remained
operable. The inspector reviewed Operating Procedure 1104-4," Decay Heat
Removal System," Revision 107 to verify that the revised system operation would
limit the flow through the drop line and eliminate the potential of reverse flow
through the reactor building sump screens during post-accident conditions. The
inspector also reviewed safety evaluation PCR 1-05-97-0519, dated October 17,
1997, and verified that these changes did not require a technical specification
l change or result in an unreviewed safety question.
1
c. Conclusions
The licensee properly documented that there was no safety concern associated with
reverse flow through the RB sump screens.
E8.20 (CLOSED) URI 50-289/96-201-20.and eel 97-070-07014.Testina of Molded Case
Circuit Breakers
i
The AE team identified that appropriate procedures had not been followed for
testing frequency of the breakers in MCCs 1 A ES ESF VENT and 1B ES ESF VENT.
, Following further reviews, the NRC identified this issue as a violation of 10 CFR 50,
'
Appendix B, Criterion XI, Test Control, as documented in a Notice of Violation and
l Imposition of Civil Penalties, dated October 8,1997. The licensee's corrective
I
actions to address this issue were reviewed, found appropriate and documented in
section M8.1 of NRC Inspection Report 50-289/98-02. These items were closed in
that report.
l
l
,
l
l
l
!
_ .---__ _ ._ ~ . - . _ ._
= ,
25
E8.21 (CLOSED) URI 50-289/96-201-21.and eel 97-070-01033. Static Head Correction
for BWST Level Transmitter
a. Insoection Scope
The AE team identified a concern involving the static head correction for BWST
level instrumentation. Specifically, the team questioned the bases for the static
head corrections for differences in elevation between the bottom of the BWST and
the level transmitters in data sheets El-1 through 4 for level transmitters DH-CT-
808 and DH-LT-209 in surveillance procedure 1302-5.19," Borated Water Storage
Tank Level Indicator," Revision 18. Subsequent NRC reviews determined that this
inadequacy constituted a violation of 10 CFR 50, Appendix B, Criterion Ill, Design
Control, as documented in a Notice of Violation and imposition of Civil Penalties,
dated October 8,1997.
b. Observations and Findinas
In response to this violation, the licensee stated that they had used a proper head
correction that was measured for these level transmitters (DH-LT-0808 and -0809)
The field sketches in question, included as part of the engineering evaluations (EER
88-070-E and EER 88-060-E), indicated the actual measured head correction. Also,
the head correction information used in surveillance procedure 132-5.19, was
obtained from these engineering evaluations. While tha EERs were reviewed and
approved, the field sketches prepared as part of the EER were not incorporated into
design documents which receive a separate design verification or design check.
The licensee took the following corrective actions:
l
- Updated borated water storage tank level transmitters drawings 308810 and
308920 on February 17,1997 to show the transmitters and BWST
elevations that form the basis for the head correction of each transmitter.
- Updated the GMS2 database to reference the change document for easy
reference change document retrieval.
- Updated BWST surveillance procedure, SP 1302-5.19, on February 5,1997,
to reflect the conservative values and reference drawings.
- lssued a new loop error calculation for BWST levelinstrument loops (C-
1101-212-E510-057),that referenced the revised P&lD drawings for the
appropriate head corrections.
- . - . . . . . - . - - - - - ~ - . - - - . - - - - - - - - . _ . - . . . -
.. ,
+
26
The inspector reviewed the revised documents and found them acceptable. The
level transmitters drawings,308810 and 308920 indicated the proper instrument '
level elevations. To preclude similar concern, the licensee's long term corrective
action (as documented in response to the NRC's 50.54(f) letter dated October 9,
4
1997), stated that they plan to update the setpoint basis program and, in that -
program, ensure that EERs which involve setpoints would receiva a design basis
verification that will check design inputs.' Through discussion with the licensee
staff, the inspector determined that this program was progressing as established
,
and is scheduled to be completed by February 1999. ,
l
c.- Conclusions
The licensee adequately addressed the concern with static head correction for
'
BWST level instrumentation. Appropriate corrective actions were taken to address
the resulting violation and long term actions were being properly tracked.
l
~ E8.22 (CLOSED)IFl 50-289/96-201-22 BWST Level Instrument Drift
i i
a. Insoection Scoos
'
The A/E team inspection identified that the licensee was inconsistent in treating the
- instrument drift values in determining the loop errors in their BWST level instrument i
l calculations. In addition, the licensee was_ unable to provide adequate bases to ,
support the instrument drift differences used in these documents.
{ l
b. Observations and Findinas ,
'
.
To resolve these inconsistencies involving the assumed instrument drift values in'
. BWST level calculations and other documents, the licensee developed a new -
calculation C-1101-212-E510-057,"TMI BWST Level Loop Accuracy-DH-LT-0808, l
2
DH-LT-0809, DH-DPS-0914, on August 19,1997. The licensee consolidated all the
different bases applicable to BWSTinstruments information fragmented in other
- - documents into this calculation to eliminate the inconsistencies. !
!
- The inspector's review of the new calculation indicated that the licensee had used j
l the guidance of the existing GPUN Nucl ear Engineering Standard ES-002, ;
Revision 4," Instrument Error Calcula':lon and Setpoint Determination" as bases to
t . develop the new calculation. The inspector noted that the drift values used for the
B'NST devices (Rosemont transmitter, Foxboro modules, Bailey indicator, Dixon !
[ ' indicator and computer input) were consistent with vendor specifications, in cases ,
where drift specification was not available from the vendor, such as Westinghouse ;
- indicator and Barton differential pressure indicating switches, historical performance !
t data were used. The inspector's review of the as left tolerances (errors) of BWST !
J' level transmitters DH-LT-0808 and 0809 indicated no concern. All supporting bases
!. were clearly defined, and appropriately shown in this calculation.
3
4
i
!
,
. - - , . , , . . , , , - - - . , . . . , ,, - -. . - , -
__ -. . . _ . - _ _ . _ _ _ ._... - _ _ _. _ _ ._ _ _ _ _ . _- _ _ _ _ . ,
-
e
'
i
l
27 i
c. Conclusions
2
Inconsistencies identified in BWST level instrument documents were resolved and
appropriate actions were taken to eliminate uncertainties in use'of instrument drift
values,
j
'E8.23 (Closed) URI 50-289/96-201-23. Selection of BWST Low Level Alarm for' Operator
Action i
I
a. Insoection Scone ' i
The AE team was concerned that critical operations were being initiated based on !
non safety-related alarms. Specifically, the team found discrepancies with operator- l
actions required in response to a low BWST level alarm from level switch DH-DPS- j
914. The switch was listed as a category 1 device while the control room ;
annunciator panel was non safety-related.
b. Observations and Findinas ,
i
The licensee had implemented temporary procedure changes in December,1996, i
requiring operators to open valves DH-V-6A/B on a low BWST level alarm from level i
switch DH-DPS-914. Per GMS 2 database, this level switch was listed as a i
Category 1 device, but it interfaced with the non safety-related main control room
annunciator panel. The panel was powered from a vital power source. To provide
redundancy for the DH-DPS-914 alarm, the licensee proposed the use of plant
,
computer alarm A0486 which had an input from DH-LT-0809 instrument loop. This i
'
loop was safety-related, however, the plant computer was not. l
I
Also, operators initiated the switch over of emergency core cooling system (ECCS) l
pump suction to the reactor building sump from the BWST based on control room I
overhead annunciator alarms. The inspector noted that the licensee's standard i
operator direction is to use redundant instrumentation where available which i
includes safety grade console indicators. To assure adherence to the above I
standard direction, the licensee revised the abnormal transient procedures (ATP
1210-6 and 10), to include the specific use of console indicators for the initiation of i
ECCS suction switch over. In addition, training was provided to the operators as
-
documented in the training material and established operator re-qualification !
schedule of cycle 97.
-
\'
The inspector verified that appropriate instructions to accomplish the switch over to
-
reactor building suction in the above procedures were incorporated, The inspector j
also verified that operators were verifying the BWST level indicator and low level
'
alarms for this process as indicated on the revised procedure. :
2- 1
i I
, !
.
'
'
I
!
.. _ . , _ _ _ __. -
, . _ ___
,. _ . _ . _. .. . . _ . . . . ~ . _ . _ . . . _ _ _ _ _ _ . _ -_ . _ _ . _ _ _ _ . -
.. ,
..
'
28
c. Conclusions
,. The discrepancy associated with the use of non safety related annunciation to
accomplish safety related actions by operators was appropriately resolved. Good
procedural guidance and training were provided.
l
E8.24 (CLOSED)IFl 50-289/96-201-24. Undocumented Modification
ac Inspection Scope
The AE team observed a hose and metal tubing attached to the open end of
,
'
discharge relief BS-V-63B. The concern was that the attachment could adversely
affect the design function of the relief because it could create a back pressure on
the relief valve or fluid build up on the discharge that could change the lift setpoint.
b. Observations and Findinas
The licensee removed the attachment from the relief ' valve and initiated actions to l
' determine the reason for the inadequacy. It was identified that personnel had not !
properly adhered to the controls of administrative procedure,1000-ADM-7350.05, I
" Configuration Changes." The procedure required that configuration changes be i
. documented and reviewed by engineering. The procedure also states that the l
activity of temporary modifications are to be controlled by AP 1013, " Temporary '
,
Modifications and Bypass of Safety Functions." Step 2.5.3.2 of AP 1013 requires
!
the restoration of component to its normal configuration prior to return to service.
.
The installation or failure to properly remove the temporary equipment was contrary
l to the requirements of the procedures.
!-
L As part of the licensee's actions taken to correct this discrepancy, system
!~ ' engineers, who routinely perform system walkdowns, were informed of the {
circumstances surrounding this event, including the consequences and corrective 1
l. actions, and management expectations. Supervisors were required to review the
l event with their shift to heighten personnel awareness to the requirements l
associated with relief valves and attachments to valves in general. l
l
'
The inspector verified that this review had been accomplished with all five crews as {
documented in operations form, Capture No.96-357. The inspector determined
that the consequence of this event was minimal. Although the condition was -
- . prohibited by the ASME code, there was no normal or accident condition that would j
'
- cause the 75 psig relief valve to lift. Therefore, this minor procedure violation was -
not subject to formal enforcement action. l
l- c. Conclusions'
!
'
The actions taken to correct the improper attachment of a hose and metal tubing to
- the discharge of relief valve BS-V-63B were adequate. This minor violation was
not subject to formal enforcement action.
J
(
l
, ,, -
.-. _ -_ __. . - _ - _._- - .. .
. ,
29
E8.25 (CLOSED) URI 50-289/96-201-25.and eel 97-0702Q 0.60M.BWST Level Transmitter
Enclosure and Heat Tracina
a. Insoection Scoce
The AE team identified a concern involving BWST level transmitter DH-LT-808. The
team found that the cover plate of the enclosure for the instrument was missing,
thereby raising the question of adequacy of the required instrument protective
barrier. Subsequent NRC reviews determined that this was a violation of 10 CFR
50, Appendix B, Criterion V, as documented in a Notice of Violation and imposition
of Civil Penalties, dated October 8,1997.
b. Observations and Findinos
The licensee attributed the cause of BWST level transmitter DH-LT-808 not being
found intact to the lack of adequate instructions in surveillance procedure 1302-
5.19, " Borated Water Storage Tank Level Indicator." The procedure did not
specifically require that the enclosure cover be replaced. Also, in the case of the
level transmitter as-found heat tracing configuration not being consistent with
applicable procedure and vendor instructions, the licensee determined that the
cause of this was due to incorrect installation in 1988, and possibly due to a lack of
experience by the installers.
The licensee's corrective actions taken to address these issues included: (1)
reinstallation of the cover plate as required; (2) updating the surveillance procedure,
1302-5.19,to include additional instructions to ensure that the enclosure cover is
closed after completion of the surveillance; and (3) correcting the level transmitter
The inspector conducted a walkdown of the accessible areas of the BWST and
noted that the cover plate of level transmitter DH-LT-808 was reinstalled properly.
The inspector reviewed the surveillance procedure 1302-5.19,and verified that the
licensee had added the appropriate instructions in the revised procedure to ensure
that the enclosure cover would be closed after completion of the surveillance. The
inspector also verified by reviewing Work Request No. 785993, completed on
June 4,1997, that the heat trace issue was appropriately corrected.
c. Conclusions
Appropriate actions were implemented to ensure that the BWST level transmitter
enclosure and heat tracing were maintained properly.
_ _ _ _ _ . _ _
-. ,
,
!
p 30
l
-
E8.261 CLOSED) URI 50-289/96-201-26.FSAR Discreoar.cies
i
a. Insoection Scope
l' l
l The AE team identified some discrepancies in sections of the FSAR regarding the
. decay heat removal system (DHRS). While the discrepancies were minor in nature,
L the team was concerned with the fact that they had not been corrected and the
l~ FSAR updated tu assure that the information included in the FSAR contained the
latest material as required by 10 CFR 50.71(e).
b. Observations and Findinas
l
in subsequent follow up, the NRC determined that although a violation of 10 CFR l
l 50.71(e), these discrepancies were of minor significance and the issue was
therefore identified as a non-cited violation in a letter dated October 8,1997 to the
'
Licensee.
l During this inspection, the inspector reviewed the licensee's actions taken to !
L correct the identified discrepancies as well as those taken to prevent further FSAR
L discrepancies. The inspector verified *: hat all the discrepancies were corrected in ;
l the April 1998 FSAR update. In a broader sense, the licensee established FSAR I
l
'
section owners to improve the quality of the FSAR and transferred the ownership of l
.the FSAR to the engineering department. !
c. . Conclusions -
l The licensee adequately resolved the FSAR discrepancies identified by the AE team.
l
In addition, the ongoing efforts by the licensee to improve the overall quality of the
( FSAR appeared appropriate.
l
E8.27 (CLOSED)IFl 50-289/96-201-27. Definition of Sinale Active Failure
l- The AE team raised a question involving the licensee's definition of " single active
failure" as it involves check valves. Section 4.1.3.7 of TMI-1's decay heat removal
system design basis document, SBDB-T1-212 states that the definition of active
component has been interpreted to exclude self actuating components for which .
there is adequate positive force to assure they function (i.e., check valves). !
Therefore, the licensee's position was that check valves were not considered to be l
" active" components.
, The NRC reviewed the licensee's position and found it to be consistent with NRC I
practice and with the licensing basis at TMI-1. Check valves, except for those in ,
'
containment isolation systems, have been treated as passive components. )
Therefore, this item is closed.
'
- !
'
i
i. ,
! !
V \
L
,
.c ,. .-- . . . . _ , . -. - -
_ _ _ _ . _ - .__ . _ _ _ .. - _ ._ _._. _ ._. _ _ __._._ _ .__.
.. ,
2
i 31
- E8.28 (CLOSED) URI 50-289/96-201-28,and eel 97-070-10153,10163,10173, and
t 0183, Control of Calculations
.
- a.- Inspection Scone
i
- 1 The AE team identified situations where safety-related calculations where not
controlled in accordance with the licensee's procedures. Following further NRC
- review, this issue was identified as a violation of 10 CFR 50, Appendix B, Criterion j
,
Ill, Design Control, as documented in a Notice of Violation and imposition of Civil
Penalties, dated October 8,1997.
b. Observations and Findinas
The licensee initiated QDR 97-2001 to document the concerns associated with the.
control of calculations. The licensee conducted a root cause evaluation and
,
identified the causes of the problem to be lack of familiarity / inadequate training on
..the requirement of procedures and lack of management enforcement to fully
implement procedural requirements. To correct this problem, the licensee initiated I
, several actions including correcting the specific probicms identified by the AE team
d
and making programmatic enhancements to prevent recurrence,
The inspector verified that the issues specified in the AE inspection report (96-201)
had been corrected as follows:
- * The calculation buried in memorandum 5310-92-366(12/22/96)," Evaluation
of TMI-1 HPl SSFI Observation No. 211-10" was incorporated into l
l calculation No. C-1101-211-E630-080. !
- The analysis for not leak testing check valves DH-V 14A and B as part of the
- IST program was nullified and the valves, including DH-V-5A and B, were
i
incorporated into the IST Program and tested satisfactorily during the cycle
'
12 refueling outage.
- TDR No. 836, " Evaluation of Loading for the Emergency Diesel Generator
and Engineered Safeguards (ES) Buses, was incorporated into design !
calculation C-1101-741-E510-005.
1
- TDR No. 995, " Voltage Drop Study for Degraded Grid Condition," was j
incorporated into design calculation C-1101-700-E510-010. !
" Calibration for DH-DPS-914," were converted into calculation C-1101-212- '
E510-057. j
i
.
l
_ _ _._ _ _ _ . _ _ _ _ . . _ . _ _ _ . _ . _ _ _ . _ _ _ . _ _ - _ _. _ _.. _ _ _ .. _ _ _
-, ,
32
As further corrective actions, the calculation process was enhanced with the
implementation of procedure E-OO6, " Calculations." In addition, the licensee did the
following: reviewed design basis calculations for key design basis parameters and
implemented the calculations into the EP-OO6 process; reviewed a sample of TDRs,
Topical Reports (TRs) for buried calculations, that needed to be incorporated into the
EP-OO6 process; and reviewed design calculations for FSAR Chapter 14 and other
selected systems. The inspector reviewed EP-OO6, Calculations, Revision 7, and
found it detailed. It is applicable to both TMl and Oyster Creek, and establishes the
requirements, methods and responsibilities for the documentation reviews, and
control of calculations generated by engineering. It contained information that
would ensure that calculations are performed, reviewed, and approved properly.
The inspector found the licensee's actions appropriate.
c. Conclusions
(
The inspector determined that the actions taken and those ongoing to identify and t
control analysis / calculations were adequate. The enhanced procedure, EP-OO6,
" Calculations," was comprehensive and contained good guidance for controlling
calculations.
V. Manaaement Meetinas
X1 Exit Meeting Summary
On August 27,1998, the inspectors presented a summary of the inspection findings to the
licensee at an exit meeting held at the site. Discussions with GPU Nuclear staff were held
subsequent to the exit meeting to confirm certain details associated with the issues
described in section E8.4. The list of persons contacted, below, reflects the participants at
the meeting, including those who participated via telephone fror.1 GPU Nuclear's corporate
office.
. }
.
t
7 1
E l
4
1
,
- - - -- , - . - - - - - ., - . . - - --
e ,,
I
l
33
LIST OF PERSONS CONTACTED
Licensee
A. Rone,VP - Engineering
J. Langenbach, VP - TMI-
. R. McGoey, Director - Engineering Support (Acting) !
- P. Walsh, Director - Equipment Reliability I
D. Skillman, Director - Configuration Control
- S. Wilkerson, Manager - Systems Engineering
- C. Hartman, Manager - Systems Engineering
' A. Asarpota, Manager - Modifications
J. Schork, Manager - Nuclear Safety
' M. Kapil, Manager - Electric Power & Instrumentation
.
. N. Trikouros, Manager - Safety & Risk Analysis
E! Fuhrer, Nuclear Safety and Compliance Committee
B. Knight, Licensing Engineer
LA.' Miller, Licensing Engineer
B. McSorley, Systems Engineer
INSPECTION PROCEDURES USED
IP 92903 Follow' up - Engineering j
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened 50-289
98-06-01 VIO Makeup System Pumps Suction Cross Connect (E8.4)
Closed 50-289 '
l
96-201-01 IFl Letdown line Break in the Auxiliary Building (E8.1)
96-201-02 IFl Evaluation of Simultaneous Start of MU&P Pump MU-P-1C and Suction
and Discharge Valves (E8.2)
96-201-03 IFl. Evaluation of Gas Accumulation in Suction Piping for Makeup Pump
.-
MU-P-1 C (E8.3)
96-201-04 URI Adequacy of Makeup Tank Pressure / Level Curves (E8.4)
'96-201-05 URI Design Basis Valve Stroke Times in Surveillance Procedure (E8.5)
-96-201-06 IFl Consequences of Failure of Auxiliary Steam Piping (E8.6)
96-201-07 IFl Loss of Pressure in MU&P Tank due to Letdown Line Break (E8.7)
'96-201-08 IFl M&UP Pump NPSH When Taking Suction From BWST or Makeup Tank
(E8.8)
, _ -l
- _ _ _ . . _ . _ - . . ~ _ _ . . _ _ . _ - . _ . _ _ _ _ _. _ _ ___ _ _ __
- ,
l.
34 !
96-201-09 IFl incomplete DC System Voltage Drop Calculations (E8.9)
96-201-10 URI Alignment of MU-V-18 DC Power Supply (E8.10)
96-201-11 UR Makeup Tank Level Instrument Loop Tolerances (E8.11)
!
96-201-12 URI FSAR Discrepancies (E8.12)
96-201-13 URI Adequacy of BWST Setpoint for DHRS Pump Switch over to RB Sump
(E8.13)
96-201-14 URI eel 97-070-01093, Adequacy of Safety Evaluation for an FSAR
Change Crediting Containment Over pressure for LPI Pumps NPSH (E8.14)
96-201-15 IFl Technical Specification Discrepancy (E8.15)
96-201-16 URI Discrepancy Between FSAR and Technical Specifications Regarding
DHRS Leakage (E8.16)
96-201-17a URI Leak Testing of DHRS Pump Suction Check Valves (E8.17a)
l 96-201-17b URI Leak Testing of DHRS Pump Discharge Check Valves (E8.17b) l
96 201-17c URI inspection of DHRS Pump Vault Floor Drain Check Valves (E8.17c)
l
96-201-18 URI Timeliness of Action on SSFl Open items (E8.18) !
l 96-201-19 IFl Evaluation of Reactor Building Sump Screen for Reverse Flow (E8.19) t
l
'
'96-201-20 URI Testing of Molded Case Circuit Breakers (E8.20)
96-201-21 URI Static Head Correction for BWST Level Transmitter (E8.21) ;
96-201-22 IFl BWST Level instrument Drift (E8.22) '
96-201-23 URI Selection of BWST Low Level Alarm for Operator Action (E8.23)
96-201-24 IFl Undocumented Modification (E8.24)
96-201-25 URI BWST Level Transmitter Enclosure and Heat Tracing (E8.25)
l 96-201-26 URI FSAR Discrepancies (E8.26)
l 96-201-27 IFl Definition of Single Active Failure (E8.27)
l 96-201-28 URI Control of Calculations (E8.28)
97-070-01023 eel Adequacy of Makeup Tank Pressure / Level Curves (E8.4)
97-070-01013 eel Adequacy of BWST Setpoint for DHRS Pump Switch over I
! 97-070-01033 eel Static Head Correction for BWST Level Transmitter (E8.21) f
l
97-070-01093 eel Adequacy of Safety Evaluation for an FSAR Change Crediting
Containment Over pressure for LPI Pumps NPSH (E8.14)
l-
97-070-03013 eel Timeliness of Action on SSFl Open Items (E8.18) -
97-070-06014 eel BWST Level Transmitter Enclosure and Heat Tracing (E8.25)
97-070-07014 eel Testing of Molded Case Circuit Breakers (E8.20)
97-070-08014 eel Leak Testing of DHRS Pump Discharge Check Valves (E8.17b)
i 97-070-08014 eel Inspection of DHRS Pump Vault Floor Drain Check Valves (E8.17c) ;
97-070-10143 eel Leak Testing of DHRS Pump Suction Check Valves (E8.17a)
i
!
- - . . - . - .-__ - _ ~_ - -
.
-.u. . , ,. - - - - - , - - - - .. _ ~ . - . - - . . . - . - _ . - . . . .
. . .-
l
.
35
LIST OF ACRONYMS USED
AE- Architect Engineering
BS Reactor Building Spray (ECCS)
BWST . . Borated Water Storage Tank
' CAP Corrective Action Process
CFR' Code of Federal Regulations
-DBD ' Design Basis Documents
.DC Decay Heat Closed
DH Decay Heat Removal (ECCS)
DHRS Decay Heat Removal System
DR Decay Heat River (ECCS)
EDG Emergency Diesel Generator
eel Escalated Enforcement item
ESF Engineered Safety Feature
EQ Environmental Qualification
FSAR Final Safety Analysis Report
GL' Generic Letter i
GPUN GPU Nuclear I
HPl High Pressure injection
IFl Inspection Follow up Item
LER Licensee Event Report
'-
LOCA Loss of Coolant accident
- LPl . Low Pressure injection
MU ' Makeup
MUT ~ Makeup Tank
, NCV- Non-Cited Violation
NRC- Nuclear Regulatory Commission
PRG Plant Review Group
SSFI Safety System Functional inspection
SWSOPi Service Water System Operational Performance Inspection
TMI Three Mile Island l
TS Technical Specification l
UFSAR Updated Final Safety Analysis Report )
URI Unresolved item
USQ- Unreviewed Safety Question
VIO Violation -
o
.
'
y w