A09604, Submits Response to Review of RI-91-A-0082 Issues Concerning Plant Activities
Text
. _ _ _ _ _ _ _ _ _ _ _ _ - - _ - _ - _. _ _ _ _ _ - - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _. _.
l O
b Geretal Off aces
- Seuen Otteet. Berlin. Connecticut ~
' IC$II.Yh~
Ei P O. DOX r70 7.7[17[.'((,0.'
H ART F ORD. CONNECTICUT D61 10270 g
J (Poh 6%4000 w e wi e==um
)
August 9, 1991 r
Docket No. 50-336
)
A09604 Mr. Charles V. Hehl, Director Division of Reactor Projects U.S. Nuclear Regulatory Commission j
Region I 475 Allendale Road King of Prussia, Pennsylvania 19406 Dear Mr. Hehl Millstone Nucloar Power Station. Unit No. 2 RI-91-A-0082 Ve have completed our reviev of the identitled issues concerning activities at Hillstone Station.
As requested in your transmittal letter, our response does not contain any personal privacy, proprietary, or safeguards information.
The material contained in this response may be released to the' public and placed in the NRC Public Document Room at your discretion.
The NRC letter and our response have received controlled and limited distribution on a "need to knov* basis during the preparation of this response.
ISSUE 1:
The viring diagrams involving Reactor Coolant Pump RTD circuits have not been updated follnving modifications made under a PDCR to replace RTD circuit knife switches with Veidmuller Test Blocks.
Drawing No.
25203-31069, Sheet 5 Revision 3, dated August-29, 1.989, does not reflect the change for at least 4 RTD circuits (TCD, TCC, TCA. TCB). The instrument loop diagrams (Draving No. 25203-28500, Sheets 140 & 146) show the Veidmuller Test Blocks. Also, in Drawing No. 25203-31069, sheet 5, the jumpers. shown between cable lead 1 and the cable shield ground on the loop diagrams are nct shovi..
In addition, access to the GRITS system, to verify the.. latest drawing revisions, is restricted in that personal access codes are only valid for 30 days.
t 9110310005 911011 PDR ADOCK 05000336;-
P.
Mr. Charles V. Hehl, Director U. S. Nuclear Regulatory Commission A09604/Page 2 August 9, 1991 Request 1:
Pleare discuss the validity of the above assertions.
If dise.repancies are found, please assess the significance of the discrepancies with the respect to plant operation and safety and discuss any actions taken or planned to correct these discrepancies.
Por clarity, our response to this issue is segregated into tvo parts.
Part
/. addresses the drawing accuracy portion of the issue and Part B addresses the question relative to GRITS access.
PART A
Background:
PDCR 2-15-86, completed in December 1986, replaced Meter Device Co. kr.if e switches with Veidmuller Inc. Test Blocks.
As a result of the PDCR, 320 instrument loops were modified which required 330 draving changes. Draving 25203-31069 Sheet 5 was not changed at that time and, therefore, was not updated.
Draving 25203-39045, Sheet SSB includes all of the information of Draving 25203-31069 Sheet 5 plus internal cabinet viring.
Draving 25203-39045 Sheet 55B vas updated at the time of the change and therefore does shov Veldmuller Test Blocks.
Response
The assertion that Draving 15203-31069 Sheet 5 was not upgraded at the time of the PDCR implementation is valid.
This ves the result of an isolated oversight aM is not indicative of a program deficiency.
Draving 25203-31069 shvet 5 is being changed to show the Veldmuller Test Blocks and jumper configuration under Draving Change Request DCR H2-S-1216-89.
PART B
Background:
Each individuti vith a need to access the GRITS system has been assigned a User Identification number by the Inforr tion Resources Group at Northeast Utilities' corporate offices.
Every 30 days individuals with access to this program vill be prompted by the computer to change their passwords.
The computer,is programmed to remind users and provides on-screen instructions on hov to change passwords. The computer is also programmed to provide a space where the user vill specify a new password.
If the terminal has not been accessed within 30 days, access is not lost.
In this case, the user must update his password prior to accessing the GRITS program. It system difficulties are encountered, an IRG HELP phone line is available (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day) as is department assistance.
Mr. Charles V. Hehl, Director U. S. Nuclear Regulatory Commission A09604/Page 3 August 9, 1991
Response
The user ID and password system is designed to provide the necessary level of security along with an appropriate level of ease of use for the person using the system. Adequate support fo* infrequent users is alsi provided.
ISSUE 2:
The Steam Generator No.
2 mid-loop instrumentation (L-122 ) vas not
" operable" durit,g drain-down for tube inspections on May 2, 1991.
GEM svitches were found to be " frozen" on in place. In addition. L-112 had an electronic noise problem caused by an improperly installed jumper.
Thus licensee commitment that two monitors be operable during drain-down condition vns not being met.
Request 2:
Please discuss the validity of the above assertions.
If any discrepant conditions are identified, please discuss their significance with respect to plant operation and safety during Steam Generator No.
2 drain-dovn evolution.
Also please discuss any actions taken or planned to correct these deficiencies.
Response
For additional clarity, this issue has been segregated into four sections.
A.
The Steam Generator No. 2 mid-loop instrumentation (L-122) was not
" operable" during drain-dovn for tube inspections on May 2, 1991.
B.
GEM switches were found to be
- frozen" in place.
C.
L-112 had an electron!
noise problem caused by an improperly installed jumper.
D.
The licensee commitment that two monitors be o,erable during drain-dovn condition was not being met.
Background:
A.
During the April / hay 1991 Steam Generator shutdovn, the Vestinghouse Ultrasonic level measuring system (L-122) vas not operable during drain-dovn for tube inspection.
L-122 was procedurally deleted from use on April 24, 1991 (Procedure Change #3 to Operations Procedure OP 2301E, Rev.
15).
During the April /May 1991 shutdown, Vestinghouse
- provided, installed, and tested enhanced design transducers and software.
Testing of the system continues until reliable system operation is achieved.
B.
The GEM level indicator (LG-112) uses a floating magnet to posi' tion
" flags" that provide a visual indication of hot leg level. These flags are monitored by closed circuit TV in the Control Room.
During the
l Mr. Charles V. Hehl, Director 1
U. S. Nuclear Regulatory Commission A09604/Page 4 i
August 9, 1991 j
unplanned April /May 1991 outage, previously identifi9d problems with f
the proper response of the flags to the magnet vare investigated.
i These problems were attributed to factory mismarked replacement flag assemblies and vere corrected.
The assembly was then tested and demonstrated as acceptable performance.
1 Subsequent to this activity, during mid-loop conditions during the April /May 1991 outage, the response of the GEM level indicator was observed
'o not change during very small
(<3/4"), and slow changes in RCS leve: at the +4.5 inch level. Troubleshooting identified that a very light tapping on the side of the GEM standpipe was sufficient to free what was suspected to be a stuck float. Float sticking vas not userved during post installation testing performed during the 1990
- efueling outage or during previous testing prior to placing the system in service. The vendor of the system had reviewed this problem. They have suggested the replacement of the existing GEM standnipe that contains an internal guide rod with a new design that elit nates the potential for guide rod binding. INECO intends to obtain and replace the existing assembly with the new design in the future.
C.
LT-112 is a level sensor that generates an analog signal representing the liquid level in a standpipe that is connected to the RCS hot leg.
The original design of the system contained an optional electronic lead circuit that was intended to improve the response time of the system during reduction in level. During the system installation and testing during the 1990 refueling outase, unacceptable performance was noted and this feature was defeated by installing a jumper.
During the April /May 1991 steam generator shutdown, unexplained bias and lov frequency output indication variations were observed. Upon further testing and consultation with the manufacturer it was determined that the location of the jumper did not completely eliminate the interference cf the lead circuit.
The jumper placement was corrected and the system response stabilized.
The bias errors observed during the January unnlanned outage vere attributed to inadequately-sized head vent tubing, and vere abserved only during fill-up or drain-down evolutions.
Larger tubing was installed during the April /May 1991 outage to correct the head vent restriction. A calibration check on May 3, 1991 of the FCI electronics (L-112) provided results very close to those obtained during preoperational testing, and factory acceptance testing at the FCI factory prior to shipment.
The lov frequency noise response characteristic of the system and the bias observed during the June 1991 outage requires additional monitoring and avaluation for appropriate corrective action.
Responses l
A.
L-122 vas procedurally deleted from use on April 24, 1991, and l
therefore was not required for drain-down during the April /May 1991 shutdown.
Troubleshooting efforts were continued in a priority basis l
R n
?
?
t Mr. Charles V. Behl, Director j
U. S. Nuclear Regulatory Commission A09604/Page 5 August 9, 1991 to restore the indication. NNECO and the vendor are continuing efforts to resolve problems associated with the application of ultrasonic technology in this application.
B.
The GEHs sensor was not found " frozen" in place as asserted.
Poor responne to small slow changes in level was noted and investigated.
NNECO is planning design improvements that vill improve the sensitivity of the indicator.
C.
The LT-122 jumpcr placement was corrected. This error did not affect the operability of the indication.
D.
Operations Procedure OP 2301E requires two operable level indicators for drain-dovn activities.
At all times while draining to reduced inventory conditions, at least two level indication systems were in operation.
These systems satisfied tne level monitoring requirements that vere in effect.
ISSUE 3:
Pressure indicating instrument (PI 6350 A/B and P1 6351 A/B) and mountings for Service Vater (SV) supply to Emergency Diesel Generators (EDG) are not seismically mounted. Any kind of shock would be sufficient to knock the gauge ac4 valve off the strainer. Additionally, the location of the taps as shown on the P&ID apparently does not coincide with the actual tap locations.
Request 3:
Please discuss the valfdity of the above assertions. If the assertions are
Please provide any actions taken or planned to ensure that seismic requirements for these instruments are being met.
Background:
The issue of the questionable mounting of the gauges and the draving accuracy was previously identified to management. The design was revi'eved and found acceptable for both dead veight and seismic loads. The draving vas reviewed and found to be correct.
Response
The assertions are not valid. No additional action is varranted.
ISSUE 4:
On May 3, 1991, the Unit 2 Stack Radiation Monitor (RM B132) vas inoperable as a result of being flooded with vater.
This monitor vould have been inoperable anyvay, as air flow had been isolated.
Filling and pressure testing of Steam Generator (SG) 41 vas underway during the same time
Mr. Charles V. Hehl, Director U. S. Nuclear Regulatory Commission A09604/Page 6 August 9, 1991 period.
Problems with valve line-ups for the radiation monitor and the SG testing contributed to the flooding and monitor inoperability.
Additionally, Health Physics (HP) control during removal of the water from the monitor was inadequate resulting in contamination of personnel.
Request 4:
Please discuss the validity of the above assertions. If discrepancies are confirmed, please discuss actions that vou have taken or vill take to ensure that plant procedures regarding 1 r stion monitor operation, conduct of tests, and HP activities are being used properly.
Response
On May 3,
1991, Operations personnel noted the loss of the Noble gas activity monitor and subsequently found vater coming out of the vent upstream of the No.
1 Atcospheric Dump Valve, and floving onto RM-8132.
The vent was closed to stop the water flov. The Chemistry Department was notified to take samples as required for an inoperable Stack Radmonitor.
As part of the valve line-up for this system, the operator signed for the vent valve to be in its required OPEN position.
Looking at the steam generator pressure test completed the previous veek, the required position for the vent valve had been changed by the Shift Supervisor from OPEN to CLOSED, and the test was completed successfully.
This was the desired position of the valve for the test. The line-up vas intended to be reviewed to indicate the actual desired position of the valve during operations.
Procedure Vriters Group individuals that vere involved thought that a change vould be put into the valve line-up by the operating shift and that the new revision vould follow on a normal schedule.
The change was not submitted, and the next test was completed with the valve in the incorrect position which alloved water to flov on to RM-8132, causing its failure.
The valve line-up errors were the result of the Steam Generator pressure test and have been corrected. The valve line-up for the Radiation Monitor is correct and no changes are necessary.
No personnel contamination resulted from this event. Ve have discussed this event with the operating personnel involved and identified to them the need for accurate valve line-up information at all times.
ISSUE 5:
Procedure discrepancies exist between OP 2336E and SP 2617A for the restoration of the line-up for the radiation monitor (RE-245), and its associatm sample pump.
Operators routinely fail to perform OP 9336E.
Section 5.1, Step 5.1.13 vhich is to immediately close A0V-244A/B and A0V-045 when securing from Condensate Polishing Facility discharges. This failure to follow procedures results in the sample pump to radiction monitor (RM-245) continuing to operate when the tank discharge is secured.
l Mr. Charles V. Hehl, Director U. S. Nuclear Regulatory Commission A09604/ rage 7 August 9, 1991 Request 5:
Please discuss the validity of the above assertions.
If a' discrepancies are identified concerning procedure noncompliance, please Iscuss their significance on the operation of tadiation monitor PH-245. Please discuss any corrective actions taken or planned, to ensure operators are meeting procedural and Technical Specification requirements.
Responset See belov.
ISSUE 6:
The following discrepancies have been fdentified during an evaluation of Vork Order AVO-H2-91-04411.
These dastrepancies identity continued noncompliance with procedures and poor response of operations and management to recurring problems with radiation monitor RE-245.
A.
The sample pump continues to run vhen the tank discharge stops at 15%
tank level (TK-ll).
B.
The "Lov Plov" svitch does not always see a flow conditio:a when TK-10 and TK-ll discharge pumps stop. The head of water in the pipe and tidal conditions affect the flow of vater.
C.
Operations normally rely on the 15% tank level pump trip to stop flow causing a lov flow to trip shut RE-245 discharge valve, and A0V-245.
If A0V-244A/B are shut and no lov flow conditions exists, RE-2456 sample pump vill continue to run until A0V-245 is shut.
D.
Changes to OP 2336E vere identified in 1989 to prevent the problems identified by AVO-H2-91-04411.
However, continued identified procedure noncompliance by Operations has caused repeated problems.
Request 6:
Please provide an assessment of the above discrepant conditions.
If the assertions are valid, please discuss their safety significance and effect on operation radiation monitor RE-245.
Please discuss any corrective actions that are being used to correct the problems.
Response
See below.
Issues 5 and 6 are identical to an issue raised by an employee via internal correspondence.
The responses to the issues are under development. There is an issue relating to system design which is consuming additional resources to evaluate and resolve. Ve plan to complete our evaluation and respond to both you and an employee who has raised this same issue by September 9, 1991.
_.____.___._._m____.
Mr. Charles V. Mehl, Director U. S. Nuclear Regulatory Commission A09604/Page 8 August 9, 1991 After our review and evaluation of the completed issues (Issues 1 through 4), we find that these issues did not present an indication of a compromise to nuclent safety. A valve line-up error was clearly made and it has been l;.
corrected.
Ve recognire the need to strive for a higher level of performance -in this area and ve are aggressively working towards this objec tive.
Ve appreciate the oppnrtunity to respond and explain the basis for our - actions.
Please contact my staff if there are any further.
questions on any of these matters.
Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY l
A E. J.fr 6c'r k'a ~
/
l Sen Wr Vice President I
cc:
V.
J. Raymond, Senior Resident Inspector, Hillstone Unit Nos.
1, 2, i
and 3 l
E.
C. Venringer, Chief, Projects Branch No.
4, Division of Reactor Projects 1
E. M. Kelly, Chief, Reactor Projects Section 4A 1
1 s
i I
s P
l i
,-..,..,.._m..
... ~.
n..