Text

Provides Response to Allegation RI-91-A-0077 Re Wiring Diagram for HPSI Pumps B & C Common Discharge Header Isolation Valve Found Not to Reflect Actual Plant Conditions.Related Info Encl
	ML20044C268
Person / Time
Site:	Millstone
Issue date:	08/30/1991
From:	Mroczka E; NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES SERVICE CO.
To:	Hehl C; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML16266A160	List: ML20044C268; ML20044C271;
References
	FOIA-92-162 A09702, A9702, NUDOCS 9303190364
	Download: ML20044C268 (20)
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. . m. . . . w-August 30, 1991 Docket No.30-336 A09702 RE: Employee Concerns Mr. Charles V. Hehl, Director -

Division of Reactor Projects U. S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, Pennsylvania 19406

Dear Mr. Hehl:

Millstone Nuclear Power Station, Unit No. 2 RI-91-A-0077 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 know" basis during the preparation of this response.

ISSUE:

The viring diagram for the High Pressure Safety Injection Pumps B and C ,

Common Discharge Header Isolation Valve (2 SI-654) was found not to reflect actual plant conditions. Specifically, contacts 12 and 13, designated

" spare" on the draving are energized with 120 VAC. The viring diagram evidently had not been updated as part of a modification done under Project Assignment 84-63 which had used the contacts. An additional concern exists in the fact that preventive maintenance activities have been routinely performed in the past with no one reporting voltage at contacts 12 and 13.

Request: .

Please discuss the validity of the above assertions. If any deficiencies are identified in viring diagrams and/or drawings, or in the- procedural control of preventive maintenance activities, please provide us with the-corrective actions you have taken to prevent recurrence. Please provide us with an assessment of the significance with regard to safety of any identified deficiencies.

9303190364 921217 em vois HUBBARD92-162 PDR-

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6 Mr. Charles U. Hehl, Director 1 U. S. Nuclear Regulatory Commission ;

A09702/Page 2-August 30, 1991 j

Response

i The viring diagram for the High Pressure Safety Injection Pumps "B" and "C" i Common Discharge Header Isolation Valve (M2-SI-654) is NUSCO Drawing !

No. 25203-31025, Sheet 3. The drawing being used _vas the current file ;

revision which identifies contacts 12 and 13 as " spare".

Project Assignment (PA)84-063 is an ongoing project for installing Thermal ;

Overload (TOL) alarms for all safety-related Motor Operated Valves (MOVs).

The purpose of the project is to provide a common motor control center (MCC) alarm in the Control Room in the event any MOV in that MCC develops a !

thermal overload condition. The spare auxiliary contact terminal points 12 and 13 in each MOV cubicle are being used to terminate wires from the l Control Room annunciators in support of this common alarm scheme. '

Vork has progressed to the point where the spare contacts in the M0V I cubicle have been vired to the annunciator system. Tags have been hung on !

inter-connecting vires to contacts 12 and 13 identifying the termination Terminating the wires to the annunciator system i points and the PA.

resulted in the application of the 125 VDC annunciator voltage across these ,

contacts.

The project is not complete, and the thermal alarm relays and local I indicating lamps have not been installed; therefore, the file copies of the j drawings have not been revised to show this design change. A' complete set i of drawings, which depict the intended configuration, have been issued for construction and are in the possession of appropriate groups at the plant, j including Millstone Unit No. 2 Engineering. The Generation Records !

Information Tracking System (GRITS) properly lists the status of viring i diagram 25203-31025 Sheet 3 as "open" with outstanding Design Change Request (DCR) M2-S-286-90 against PA 84-063.

t It is not necessary to have potential across contacts 12 and 13 at this ,

time in the implementation of the project. Therefore, the power leads at ;

the annunciator cabinets have been lifted, which removes the voltage from '

the " spare" terminals in the MOV MCC cubicles. !

The assertion described above is not valid and there is no significance with regard to safety. The GRITS properly identified the status of the !

drawing and the open PA and DCR associated with the proj ect , therefore, I there is no deficiency in either the viring diagram or the drawing control i system. All electricians and mechanics have received training on use of i the GRITS and have been instructed to use the system to determine the current status of drawings. Proper use of the GRITS vould have alerted a I

user to an outstanding change to the drawing and further research would ,

have revealed that contacts 12 and 13 could be energized. Therefore, when the proper draving review process is followed, there is no need to report ,

the existence of the voltage on contacts 12 and 13. Based upon the above, [

no corrective actions are needed. }

f

I w Mr. Charles V. Hehl, Director ,

i. U. S. Nuclear Regulatory Commission A09702/Page 3 August 30, 1991 This issue was previously identified to us by an employee and a vritten response was provided on a timely basis with a complete explanation of this situation, along with recommendations for future troubleshooting activities.

After our review and evaluation of this issue, we find that this issue did not present any indication of a compromise of nuclear safety. Ve appreciate the opportunity to respond and explain the basis of our actions.

Please contact my staff if there are further questions on any of these matters.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY E.JT)Roczka g Senior Vice President cc: V. J. Raymond, Senior Resident Inspector, Millstone Unit Nos. 1, 2, and 3 '

E. C. Venzinger, Chief, Projects Branch No. 4, Division of Reactor Projects E. M. Kelly, Chief, Reactor Projects Section 4A J. T. Shedlosky, NRC, Millstone Nuclear Power Station i

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, (203) 66N August 30, 1991 Docket No. 50-336 A09702 1[ . j' RE: Employee Cone s j JjA/

Mr. Charles V. Hehl, Director f (M jY' >

Division of Reactor Projects -

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Dear Mr. Behl:

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RI-91-A-0077 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 know" basis during the preparation of this response.

ISSUE:

The viring diagram for the High Pressure Safety Injection Pumps B and C Common Discharge Header Isolation Valve (2 SI-654) was found not to reflect actual plant conditions. Specifically, contacts 12 and 13, designated

" spare" on the drawing are energized with 120 VAC. The viring diagram evidently had not been updated as part of a modification done under Project Assignment 84-63 vhich had used the contacts. AnJdditional concea 'ggib/ .kib/

f /. -

exists in the fact that preventive maintenance activities have been -'

routinely performed in the past with no one reporting voltage at contacts ?

12 and 13. c>

Request:

Please discuss the validity of the above assertions. If any deficiencies are identified in viring diagrams and/or drawings, or in the procedural control of preventive maintenance activities, please provide us with, the - 7 corrective actions you have taken to prevent recurrence. Please p' rov de us afe with an assessment of the significance with regard t

~[ tty (Cd' ^ ' [y > ap I identified deficiencies.

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4 a Mr. Charles V. B;hl, Director

! U. S. Nucicar Regulatory Commission A09702/Page 2 August 30, 1991

Response

The viring diagram for the Bigh Pressure Safety Injection Pumps "B" and "C" Common Discharge Header Isolation Valve (H2-SI-654) is NUSCO Drawing No. 25203-31025, Sheet 3. The drawing being used was the current file revision which identifies contacts 12 and 13 as " spare".

Ploj ect Assignment (PA)84-063 is an ongoing project for installing Thermal Overload (TOL) alarms for all safety-related Motor Operated Valves (MOVs).

The purpose of the project is to provide a common motor control center (MCC) alarm in the Control Room in the event ant MQV in that~MCG deveJLops a thermal overload condition. The_ spare auxiliary E6ntact terminal points _12 and 13 in each_ MOV cubicle are beine used to_ terrMnate vires from the control noom-annunciators in support of this common alarm scheme.

Vork has progressed to the point where the spare contacts in the MOV cubicle have been vired to the annunciator system. Tags have been hung on inter-connecting vires to contacts 12 and 13 identifying the termination points and the PA. Terminating the vires to the annunciator system resulted in the application of the 125 VDC annunciator voltage across these contacts.

The proiect is not complete, and the thermal alarm relays and local indicating lamps have not been installed; therefore, the file copies of the drawings have not been revised to show this design change. A_gomplete sel of drawints, which depict the intended configuration, have been issued for construction ano are in the possession of approoria_te groups at the plant, Qeludinz Millstone Unit No. 2 Enrineering. The Generation Records Information Tracking System (G I properly lists the status of viring diagram 25203-31025 Sheet 3 ,R,IT)) as h open" with outstanding Design Change Request (DCR) M2-S-286-90 against PA]4f-D63.

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,AgtU M t is not necessary to have potential across contacts 12 and 13 at this V / *m Wga time in the implementation of the project. Therefore, the power leads at 14 the annunciator cabinets have been lifted, which removes the voltage from the " spare" terminals in the MOV MCC cubicles.

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, h ertion described above is not valid l and there is no significance vith regard to safety. The GRITS properly identified the status of the drayjng and the open PA and FCR associated with the proj ect , therefore, there is no deficiency in either the viring diagram or the drawing control system. All electricians and mechanics have received training on use of j

the GRITC and have been instructed to use the system to determine the

' current status of drawings. .Pyoper e use of the GRITS vould have alerted a user _to an outstanding change to tne craving and furthar_ research would have revealed that contacts 12 and 13 could,be energized. Therefore, when the proper draving reviev process is foll 'ed, there is no need to report the existence of the voltage on contacts 2 and 13. Based upon the above, no corrective actions are needed. t o

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1 i Mr. Charles U. Behl, Director

! U. S. Nuclear Regulatory Commission A09702/Page 3 ,

August 30, 1991 ,

This issue was previously identified to us by an employee and a written response was provided on a timely basis with a complete explanation of this situation, along with recommendations for future troubleshooting activities. L Af'.er our reviev and evaluation of this issue, we find that this issue did rot present any indication of a compromise of nuclear safety. Ve appreciate the opportunity to respond and explain the basis of our actions.

Please contact my staff if there are further questions on any of these .

matters.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY E. JQWoczka f Senior Vice President ec: V. J. Raymond, Senior Resident Inspector, Millstone Unit Nos. 1, 2, and 3 E. C. Venzinger, Chief, Projects Branch No. 4, Division of Reactor :

! rojects P

E. M. Kelly, Chief, Reactor Projects Section 4A J. T. Shedlosky, NRC, Millstone Nuclear Power Station I i

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. /q f .y inspector concluded there is no present operability concern with the

, p q; RPS channels of interest (RCS flow, RCP speed and zero mode bypass). The functional test should be changed to comply with IEEE 338 (1971) for the reasons stated above.

The failure to test the RPS channels as close to the sensor as practicable during the monthly functional test is a deviation from a licensee

/[f3 commitment. This is the second of two deviations identified during this p (fh

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[h inspection (50-336/90-22-03).

.,4 5.3.3 Wide Range Nuclear Instrumentation Operability - Unit 2 jp

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Inspector review of refueling activities on October 8,1990, noted that (7 '9 i reactor engineering and operations personnel were using wide range nuclear instrumentation (WRNI) channels A, B, and D, for core j

7 i' monitoring during fuel moves. Channel C was available for indication y' but was not used to meet technical specification 3.9.2 requirements.

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Ahhough channel A " spiked" periodically, it was considered by the licensee to be operable and providing an accurate indication of core

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[) conditions. It tracked fuel moves and correlated with other monitors.

Operability was demonstrated by completion of the normal surveillances.

Inspector review of a computer generated plot of the three channels for

'bjh the day shift showed stable indications for the period with the exception

,g n of two " spikes." Reactor engineering personnel respcnded to the spikes

[V by treating them as valid until proven spurious by comparison to other channels.

In addition to monitoring count rate during core alterations, data from the WRNI was used to complete 1/M plots for each core insertion.

Inspector review of the WRNI tabulated data and the 1/M plots showed that at least two channels (more often three) were always available during core alterations. The spiking problem on channel A did not preclude using the data to track core conditions during fuel moves. The inspector noted that high reactor boron concentrations (greater than 1950 ppm) resulted in low counts from all WRNI channels (in the range from 1 to 6 cps). The resulting large scatter in the data made the 1/M plots acceptable but minimally effective.

Based on the above, the inspector concluded that the technical specification requirements were being met and that core conditions were being monitored adequately by the licensee during core alterations.

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.c ws.a e w ow-HARTFORD. CONNECTICUT 06141-0270 (203) 665-5000 f1 December 21, 1990 Docket No. 50-336 A09163 Mr. E. C. Venzinger, Chief Projects Branch No. 4 Division of Reactor Projects U. S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, Pennsylvania 19406

Dear Mr. Venzinger:

Millstone Nuclear Power Station, Unit No. 2 RI-90-A-0180 arni RI-90 .' 4 0_2 3 Ve have coropleted our review of an allegation concerning activities at Millstone Unit 2 (RI-90-A-0180 and RI-90-A-0202). As requested in your transmittal letter dated November 26, 1990, 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 know" basis during-the preparation of this response.

RI-90-A-0180 --

t1: Ld LE 330 01 Issue 1.a.

Vide range Id lM hdfi N vere not operable on October 9, 1990 as required to support refueling operations because:

1) "A" channel spikes periodically. This is a long standing problem that has not been resolved, the ILC technicians have " banged on" the channel to stop the spiking.

Y osam m u, c \ oH 1 $o195

l Mr. E. C. V:nzinger, Chief j U. S. Nuclear Regulatory Commission ;

j A09163/Page 2 i l December 21, 1990

Background

1 Millstone Unit No. 2 Technical Specifications require that two of the four '

vide range flux monitor channels be operable during shutdovn and core alterations. On October 9, 1990, vide range channels A, B, and D vere ;

operable and being used to satisfy the Technical Specifications 7 requirement. Intermittent spiking of the draver indication has been a ;

long standing problem. Over the past two years, six AV0s have been '

implemented to troubleshoot ar.1 resolve this problem. Efforts to ,

determine the exact cause in the past have been inconclusive because the spiking is not repeatable on demand. Recent efforts to isolate'the fault ;

included interchanging the A drawer into the C cabinet. This has proven l successful in isolating the problem to that draver as opposed to the rest !

of the channel's components. Current plans for resolving this problem ;

include preparation of a spare drawer to allov one for one replacement and :

evaluating the need to upgrade the system to one of a design easier to maintain and more resistant to EMI. ;

Response

It is not an acceptable practice to " bang on" electronic equipment to resolve a problem. As the spiking problem with this draver is an '

intermittent one, poor electrical connections are a possible cause. ;

Movement of the draver and the components within it has been attempted in an effort to determine a causal relationship. No repeatable response has been established.

?

2) "C" channel cable has been damaged and this damage has affected readings on the channel. The channel has " low IR" readings on the cable.

Response i While replacing the channel "C" cabling, the cable outer conductor was damaged. NCR 290-110 documented this damaged condition and described the repair in the disposition details. Subsequent testing of the cable was performed under AVO M2-90-Il450 and shoved the repair to be satisfactory and the "IR" to be within the specification limit.

3) A PDCR to change out the channels continues to be open and until closed !

and signed off, the channels cannot be operable.

Background

PDCR MP2-90-072 vas written to address the replacement of vide range cable pull boxes and junction boxes. The equipment was accepted by Operations -

on November 2,1990 af ter satisfactory retest. The PDCR vas closed on November 9, 1990.

i

Mr. E. C. Venzinger, Chief U. S. Nuclear Regulatory Commission A09163/Page 3 December 21, 1990

Response

The implementation of a PDCR is controlled by a work order (AVO). The activity is authorized by Operations, performed by the appropriate work group, tested, reviewed by the appropriate engineer, and then accepted by Operations. Operations then determines the operability of the system based on the overall system status. The closecut of the PDCR document follows the closeout of the AVO document. Its status, after the AVO is accepted by Operations, does not affect the operability of the system.

Issue 1.b.

The I&C technicians have been under pressure to allow the (above) discrepant conditions to continue to exist with the channel considered operable to allow fuel alterations to occur.

Response

The intermittent spiking of the "A" vide range draver has been a frustrating problem for I6C technicians to deal with. During refueling activities, the appropriate conservative action has been taken when a problem such as EMI interference has caused any of the operable channels to be of suspect status.

These include Steps 5.5 and 5.6 of Engineering Procedure EN-21008.

During the refueling activities of the 1990 refueling outage, technical specification requirements for flux monitoring vere met. No situations associated with the vide range nuclear instruments during core alterations occurred that required specific determination of operability by PORC.

Maintenance and surveillance testing was done in accordance with PORC approved procedures. The detector, junction box, and cable replacement activities were accomplished to correct EE0 deficiencies. PORC has reviewed and approved an operability evaluation of the vide range nuclear instrumentation on December 12, 1990, PORC 12-90-192. This evaluation addressed the environmental qualification of the system as required by 10CFR50.49.

Issue 2 The "ovner" of the vide range nuclear instrument procedure, SP-2417H, was not consulted for a recent procedure change processed to support outage activities. This is contrary to I&C Department policy.

Background

ACP-0A-3.02 contains the station requirements for the reviev and approval of procedures. Procedure revisions are required to be reviewed by thre department head and by PORC. Unit 2 I&C has a department specific instruction (3.01) on department procedures to ensure that consistent, high quality procedures result from the department's efforts. Department instruction 3.01 currently includes guidance on the development of revisions. It also discusses the use

Mr.'E. C. Venzinger, Chief U. S. Nuclear Regulatory Commission A09163/Page 4 December 21, 1990 of a routing sheet by the PMMS planner as a means to coordinate revision development. The current routing sheets being issued by the PMMS planner include a flovpath of possible desired reviewers. It is up to the PMMS planner to indicate what scope and the number of reviews necessary for any given revision.

The concept of each procedure having a procedure owner was implemented in 1988 to make the procedure review and revision process more effective and efficient. Previously, procedures were not assigned individual responsibility below the department head-level. This concept has proven effective in allowing the procedure owner to be the focal point for resolving issues associated with the procedure.

The refueling outage vide range nuclear instrumentation work activities vere -;

assigned primarily to one I&C specialist. This specialist was assigned to- ;

dayshift throughout the outage. During the work activity, he found the .j' existing procedure deficient and prepared the necessary revision. The PMMS planner, with the department head's concurrence, deleted the normal practice ,

of having one of the reviews done by the procedure owner. The basis for this .I change in the normal department practice was the availability of other and !

better qualified reviewers. The procedure "ovner" was on nightshift during i this time frame and was involved in other important issues of his own. i Response }

i Revision 3 of IC-2417H was not reviewed by the procedure "ovner". Adequate review in lieu of the procedure " owner" did occur. This issue was raised by -

the procedure "ovner" and was addressed by the department head. l t

RI-90-A-0202 Is f Authorized k Order M2-90-00579 is a one-page AVO for annual pr tive i maintenance (P n various limitorque operators. A note on AVO says that l the performance of PM vill not affect EE0 boundarie ovever, ACP-2.16 Page 21, Item D states t all maintenance work o O examinations be i documented on three-page A . -l

a. Vas the one-page AVO appropria r this maintenance item? Vere there proper E0 revievs? e

Response

In 1986, Unit Maintenance reviewed PMs involving EE uipment in order !

i to determi which PMs did not affect EE0 boundaries. As sult of this evalua n, AV0s for PMs that do not affect EE0 boundaries cont the sta34 ment, " NOTE: The performance of this PM vill not affect the EE .

boundaries per R. Bonner 3/31/86". Based on.a request from Unit 2, a- l f

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gy' RECORD OF ALLEGATION PANEL DECTSTONS SITE: Mi fs > t i PANEL ATTENCEES:

ALLEGATION NO.: 47-9 I- A-oc)79 Chairman - 7. T Mnikt .

DATE: B A4 AY9d (Mtg.@2345) Branch Chief - f. (, tU e m 4 A q e r-PRIORITY: High @ Low Section Chief (AOC) - f.m. //e/f7 SAFETY SIGNIFICANCE: Yes No Unknown Others -C.u>.Kdde 2. 2. ll,w, CCNCURRENCE TO CLOSECUT: DD h SC b.T llefoly 3. S* Shad CONFIDENTIALITY GRANTED: Yes (See Allegation Receipt Report) @

k C. Oc-d 2.[. k tonied'er IS THEIR A DOL FINDING: Yes h IS CHILLING EFFECT LETTER WARRANTED: Yes No HAS CHILLING EFFECT LETTER BEEN SENT: Yes No HAS LICENSEE RESPONDED TO CHILLING EFFECT LETTER: Yes No ACTION:

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NOTES:

UNITEo STATES M

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NUCLEAR REQUt.ATORY COMMISSON MSOION I 478 ALLENDALE ROAD

- - - - KING OF PRUSSIA. PENNSYLVANIA 19406 AUG 06 E Docket Nos. 50-336 File Number RI-91-A-0079 Northeast Nuclear Energy Company ATTN: Mr. E. J. Mroczka Senior Vice President - Nuclear Engineering and Operations Group P.O. Box 270 Hartford, Connecticut 06141-0270

Dear Mr. Mroczka:

The U.S. Nuclear Regulatory Commission recently received information concerning activities at the Millstone Nuclear Power Facility, Unit 2. The details are enclosed for your review and followup.

We request that the results of your review and disposition of these matters be submitted to Region I within 30 days of the date of receipt of this letter. We request that your response contain no personal privacy, proprietary, or safeguards information so it can be released to the public and placed in the NRC Public Document Room. If necessary, such information shall be contained in a separate attachment which will be withheld from public disclosure.

The affidavit required by 10 CFR 2.790(b) must accompany your response if proprietary information is included. Please refer to file number RI-91-A-0079 when providing your response.

The enclosure to this letter should be controlled and distribution limited to personnel with a "need to know" until your investigation of the concern has been completed and reviewed by >

NRC Region I. The enclosure to this letter is considered Exempt from Public Disclosure in accordance with Title 10, Code of Federal Regulations, Part 2.790(a). However, a copy of this letter excluding the enclosure will be placed in the NRC Public Document room.

The response requested by this letter and the accompanying enclosure are not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511.

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n N:rtheast Nuclear Energy Company 2 Your cooperation with us is appreciated. We will gladly discuss any questions you have concerning this information.

Sincerely, Cha W Division of Reactor Projects

Enclosure:

(10 CFR 2.790(a) Information) cc w/o encl:

Public Document Room (PDR) local Public Document Room (LPDR)

State of Connecticut r

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bec w/ encl:

W. Raymond, SRI, Millstone Allegation File, RI-91-A-0079 turnover, 0082 (update only)

J. Stewart T. Shediosky i

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LIMITED DISTRIBUTION v - NOT FOR-PUBLIC sDISCLOSLR -

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Issue 79-1 !

Procedures being issued in the I&C department are inadequate in that acceptance criteria are !

)

i not being established for required measurements. Specifically a draft copy of procedure IC ;

2416G was provided for review and it had not incorporated several comments that were i raised on previous revisions. ' These comments included: 1) An acceptance criterion for the j output of the pulse height discriminator was not established; 2) A precaution was not added l to check the power supply output of the NLW-3 drawer if the Gammametrics power supply drops below 15 volts; The Gammametrics and NLW-3 drawers share the same power supply and the Gammametrics output acceptance criterion is 15 i 1.5 VDC while the NLW-3 output acceptance criterion is 15 i 0.0075 VDC, Therefore the Gammametrics drawer may be in >

specification while the NLW-3 is out of specification; 3) The proposed acceptance criterion !

for the discriminator bias voltage was inadequate at .9 i 1 VDC; Gammametrics l recommends 0.8 to 1.0 VDC.

In addition PORC meeting 2-89-123 authorized change No. 3 to procedure IC-2417I-1. The )

change authorized new settings for NLW-3 drawer discriminator voltage. Section 5.4 of l IC-2417I-1 should have also been changed at this time, and a discussion section on NLW-3 i i disenmmator settings should have been added. .{

i i j Request 79-1 l 4 Please discuss the validity of the above assertions. Was the procedure released for use, and if .

so, was it unusable in this field? Please state whether or not the procedural changes were )

j required to satisfy regulatory requirements, and discuss the review process for procedures and l

how comments raised during the procedure review are addressed. ;

~

l Issue 79-2 4 While troubleshooting a disabled ICC thermocouple, it was noted that a Litton-Veam i connector used to perform the troubleshooting was identical to what is installed in the ICC l system. The connector used for the troubleshooting was obtained from the NNECO {'

4 warehouse and was not EQ. Work order AWO M2-90-13287, used for the troubleshooting, did not reference procedures IC 2421C and IC 2821E, which pmvide guidance to personnel for working on Litton-Veam connectors that are EQ.. I&C department supervisors were unaware that this section of the ICC system cabling was EQ. Also the loop folders which l were being used for this work were out of date. )

I Request 79-2 l Please discuss the validity of the above assertions. If valid, please discuss actions takes to !

ensure that EQ requirements were met in this case. l 3

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- LIMITED DISTRIBUTION - NOT FOR PUBLIC DISCIDSUdE_/ !

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Issue 79-3 Recently, a PDCR which installed an audio monitoring system on the pressurizer safety valves was authorized. The audio system did not contain a spare hookup as shown on the PDCR drawing. Also the wire hookup in the PDCR showed two different setups. These problems caused the job to be delayed resulting in excessive radiation exposure of the workers.

Request 79-3 Please discuss the validity of the above assertions. If valid, please discuss the methods used to ensure that procedures are technically correct prior to performance and what preparations ,

are performed to ensure worker radiation exposure is minimized.

t r

Issue 79-4 !

On April 22,1991 the "B" and "D" battery chargers were removed from their normal power supply.12ter, a technician pe1 forming a surveillance on the "C" train Nuclear Instrument placed the instnsment into Test due to a spiking problem. 'Dtis action rendered three trains ,

of nuclear instrumentation inoperable. Since the plant was in Hot Standby at the time of the ,

test, two trains of nuclear instrumentation were required to be operable. A Plant Incident i Report (PIR) was initiated to document the occurrence but the PIR was later canceled based !

I upon an interpretation of the Technical Specification requirements.

Request 79-4 Please discuss the validity of the above assenions. Please discuss the basis for the decision j not to document the occurrence, if substantiated, with a PIR.

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HARTFORD, CONNECTICUT o6141-o270 k L J 7,$CC,E~, (203) 665 5000 September 27, 1991 Docket No. 50-336 ;

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A09768 RE: Employee Concerns i

Hr. Charles V. Hehl, Director Division of Reactor Projects t U. S. Nuclear Regulatory Commission l Region I ,

475 Allendale Road ;

King of Prussia, Pennsylvania 19406

Dear Mr. Behl:

Millstone Nuclear Power Station, Unit No. 2 RI-91-A-0079 i l

Ve have completed our review of identified 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 these responses may be released to the public and placed in the NRC Public Document Room at.your discretion. [

The NRC transmittal letter and our response have received . controlled and '

limited distribution on a "need to know" basis during the preparation of i this response. Additional time in which to respond to these issues was ;

granted by the Region I Staff in a telephone conversation on September 19, 1991.

ISSUE 79-1: ;

Procedures being issued in the I&C department are inadequate in that acceptance criteria are not being established for required measurements.

Specifically a draft copy of procedure IC 2416G was provided for review and.

i it had not incorporated several comments that vere raised on previous revisions. These comments included: 1) An acceptance criterion for the i output of the pulse height discriminator was not established;. 2) A precaution vas.not added to check the power supply output of the NLV-3 .,

drawer if the Gammametrics power supply drops below 15 volts; The !

Gammametries and NLV-3 drawers share the same power supply and the-Gammametrics output acceptance criterion is 15 + 1.5 VDC while the NLV-3 output acceptance criterion is 15 + 0.0075 VDC. Therefore the Gammametrics l drawer may be in specification while the NLV-3 is out of specification; 3) ;

The proposed acceptance criterion for the discriminator bias voltage was inadequate at .9 + 1 VDC; Gammametrics recommends 0.8 to 1.0 VDC.

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Mr. Charles U. Hthl, Director a U. S. Nuclear Regulatory Commission

[ A09768/Page 2 September 27, 1991 !

In addition PORC meeting 2-89-123 authorized change No. 3 to procedure IC 24171-1. The change authorized new settings for NLV-3 draver discriminator i voltage. Section 5.4 of IC-24171- 1 should have also been changed at this !

time, and a discussion section on NLV-3 discriminator settings should have i been added. l Request 79-1:: [

Please discuss the validity of the above assertions. Was the procedure released for use, and if so, was it unusable in this field? Please state whether or not the procedural changes were required to satisfy regulatory requirements, and discuss the review process for procedures and how comments raised during the procedure review are addressed. !

Response 79-1: I The assertion that inadequate procedures are being issued in Hillstone Unit No. 2 Instrumentation and Controls (I&C) is not valid.

The change number and Plant Operations Review Committee (PORC) meeting number stated in the assertion are for a change to the I&C Form and not the _

procedure as stated. I&C Form 24171-1, Section 5.4, was in fact changed in July 1989, to authorize new settings for the NLV-3 drawer discriminator

voltage. Discussion sections are not typically added to I&C data sheets !

and none was added in this case. Discussion sections are more appropriately included in the body of the procedure. In this case, a ,

di:,cussion section on the discriminator settings was judged to not constitute necessary information. ,

1 Ve vere previously aware of the need for revisions to the procedure at !

issue and I&C procedure IC 2416G, Vide Range Discriminator Adjustment, is currently in the revision process. The review process for procedure :

revisions includes incorporation of format changes as required by the procedure upgrade group; a review by the person responsible for the procedure, typically an instrument specialist; and independent review and !

validation activities as deemed necessary. The person responsible for the ;

procedure coordinates the resolution of comments raised through the review process. Engineering input is solicited as required to resolve issues.

The procedure is then reviewed by the department head and presented for PORC approval.

The draft revicion has incorporated many changes of both a technical and !

format nature. The copy referenced has not been issued for use in the I field. Comments are still being researched and information is still being !

incorporated. The changes being made include guidance from Gammametrics, the vendor presently responsible for support of the system. When all the existing comments have been resolved the procedure vill be re-routed for i final comments. When comments on the final draft are resolved, the l procedure revision vill be taken to PORC for review and approval. The '

procedure vill then be issued for use in the field. This procedure change is intended to enhance the use of the procedure in the field and the changes being incorporated vere not the result of any regulatory requirements.

I l

Mr. Charles U. H:hl, Director U. S. Nuclear Regulatory Commission

?'

A09768/Page 3 September 27, 1991

. ISSUE 79-2:

While troubleshooting a disabled ICC thermocouple, it was noted that a Litton-Veam connector used to perform the troubleshooting was identical to what is installed in the ICC system. The connector used for the '

troubleshooting was obtained from the NNECO varehouse and was not EQ. Work order AVO M2-90-13287, used for the troubleshooting, did not reference ,

procedures IC 2421C and IC 2821E, which provide guidance to personnel for :

vorking on Litton-Veam connectors that are E0 I&C department supervisors '

vere unaware that this section of the ICC system cabling was E0. Also the ;

loop folders ser.ch vere being used for this work were out of date. ;

Request 79-2: ;

Please discuss the validity of the above assertions. If valid, please ;

discuss actions taken to ensure that EO requirements vere met in this case.

1 Response 79-2: ,

The assertion as stated is not valid. Ve vere made aware of this issue during performance of troubleshooting activities under the Automated Vork l Order (AVO).

Troubleshooting is a logical approach to solving a problem. It is not i 1

unusual to use similar, but not qualified, equipment during troubleshooting

, because this equipment is not left installed in the system.

The original issue of the AVO did not reference the Electrical Environmental Qualification (EEQ) maintenance procedures but did contain the information that the work vas on an EEQ system. There are no special maintenance activities required to maintain the EE0 boundary of this equipment. The procedures mentioned in this assertion contain information !

on the reactor vessel head cabling removal and testing (2421C) and head area cabling support system connector assembly (2421E). The connector ;

assembly procedure (2421E) does not contain maintenance guidance for ;

testing or troubleshooting existing connectors and was not considered ;

relevant to the AVO. Procedure 2421C contains maintenance information and !

this reference vas added to the AVO job description in response to the specialist's questions during the work activity. [

t The assertion that the loop folders are out of date is not valid as there [

are no loop folders for the ICC thermocouples. !

ISSUE 79-3: ,

Recently, a PDCR vhich installed an audio monitoring system on the i pressurizer safety valves was authorized. The audio system did not ,

contain a spare hookup as shovn on the PDCR draving. Also the vire hookup in the PDCR shoved two different setups. These problems caused the job to [

be delayed resulting in excessive radiation exposure of the virkers.

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Mr. Charles V. B;hl, Director' U. S. Nuclear R:gulatory Corminsion >

f, ' A09768/Page 4 September 27, 1991 ,

. Request 79-3:

i Please discuss the validity of the above assertions. If valid, please discuss the methods used to ensure that procedures are technically correct prior to performance and what preparations are performed to ensure worker radiation exposure is minimized.

Response 79-3: ,

The assertion concerning excessive radiation exposure is not valid. The total radiation exposure for the job amounted to 0.24 manrem. The work scope accounted for in this total included construction and removal of staging to accomplish the work on the shield assemblies. The extra time in i the area that can be attributed to the confusion caused by the drawings and procedure figures is approximately 1 manhour, and approximately 0.019 manrem. While any unnecessary exposure is undesirable, this is not considered to be excessive. ,

The Acoustic Flov Valve Monitoring System (AVMS) installed at Millstone Unit No. 2 utilizes two shield assemblies as was noted in the original and latest revision of the maintenance procedure. The vendor drawings shov, and technical information states, that the system can be supplied with up .

to three charge converters in one shield assembly. The plant drawings and !

installation and maintenance procedure IC 2417T, figures and attachments. l vere developed from the vendor information and made reference to a third 4 charge converter as being there but not used. A note in the body of the procedure states that only two charge converters are used at Unit No. 2.

The "vire hookup... setup" refers to the vendor drawings which show !

different methods for different signal conditioning equipment. The PDCR i correctly referred to the proper method for the equipment installed at ,

Millstone Unit No. 2.

Ve vere made aware of this issue during vork performed under an AVO in May -

1991. The installation and maintenance procedure, IC 2417T Rev. 1, has i been changed to clarify the internal part arrangement shown in Figure 8.2 and the number of preamplifier assemblies described in Attachment 10.2 to ;

indicate that two charge converters are installed. Also NUSCO drawings l 25203-28500 sh. 193, 194, 298, 299, have been revised, via Design Change

Request (DCR) No. M2-P-015-91, to remove any reference to a " spare" charge ;

converter.

ISSUE 79-4: ;

i On April 22, 1991 the "B" and "D" battery chargers were removed from their [

normal power supply. Later, a technician performing a surveillance on the l "C" train Nuclear Instrument placed the instrument into Test due to a j spiking problem. This action rendered three trains of nuclear ;

instrumentation inoperable. .Since the plant was in Hot Standby at the time !

of the test, two trains of nuclear instrumentation vere required to be operable. A Plant Incident Report (PIR) was initiated to document the- !

occurrence but the PIR vas later canceled based upon an interpretation of ,

the Technical Specification requirements. !

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Mr. Charles V. Brhl, Director U. S. Nuclear Regulatory Co: mission 6 A09768/Page 5 September 27, 1991 j Request 79-4: i Please discuss the validity of the above assertions. Please discuss the basis for the decision not to document the occurrence, if substantiated, !

i vith a PIR.

Response 79-4: l The conditions described in the assertion are confusing and could not be substantiated as written.

The "D" battery charger supplies the turbine battery, has only one power supply, and has nothing to do with nuclear instrumentation.

There is no entry in the Shift Supervisor's log that any electrical bus alignments related to nuclear instrumentation were changed on April 22, 1991. The nuclear instruments (RPS) are povered from vital 120 vac distribution panels, which are in turn supplied by inverters povered from the vital station batteries. The station batteries each have a dedicated charger, and there is a "sving" (backup) charger. None of these systems were realigned on April 22, 1991.

The Production Maintenance Management System (PMMS) history was reviewed and no work vas found to have been performed on any battery chargers on April 22, 1991, or for a few days on either side of April 22. On April 26, maintenance was performed on the 'C' battery charger, which is the non-dedicated charger and would not have affected the operability of any DC

'ousses.

There is no entry in the Plant Incident Report (PIR) log to suggest that a PIR vas " initiated" on any related topic on April 22, 1991, and neither the Administrative Control Procedure (ACP-0A-10.01) nor NNECO practice allows for PIRs to be " canceled".

Ve were not aware of this issue as a concern prior to receipt of the NRC transmittal letter.

l After our review and evaluation of these issues, ve find that these issues did not present any indication of a compromise of nuclear safety.- Ve appreciate the opportunity to respond and explain the basis of our actions.

Please contact my staff if there are further questions on any of these matters.

Very truly yours, ,

t NORTHEAST NUCLEAR ENERGY COMPANY E. J.[iroczka f/

Senidr Vice President ec: See Page 6 t i

~

. Mr. Chnrles V. H:hl, Director

,. U. S. Nuclear Rrgulatory Cornission i A09768/Page 6 September 27, 1991 cc: V. J. Raymond, Senior Resident Inspector, Millstone Unit Nos. 1, 2, and 3 E. C. Venzinger, Chief, Projects Branch No. 4, Division of Reactor Projects E. M. Kelly, Chief, Reactor Projects Section 4A J. T. Shedlosky, IRC, Millstone Nuclear Power Station

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,, UNITED STATES -@

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Docket Number: 50-336 .

Northeast Nuclear Energy Company

-l ATTN: Mr. John F. Opeka Executive Vice President - Nuclear j Engineering and Operations Group P.O. Box 270 Hartford, Connecticut 06141-0270 '

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Dear Mr. Opeka:

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.i Thank you for informing us of the results of your ret :ews of the concerns listed in the enclosed i table. We have performed verification inspections on selected issues, find your responses generally acceptable, and plan no further actions on these issues at this time. This is not to say <

that further independent reviews of these issues will not take place in the future. You will be j

kept informed of such verification inspections and independent reviews by the normal inspection ;

report process.

l A copy of this letter as well as the referenced correspondence is being placed in the Public

~i Document Rooms and sent to the State of Connecticut. We appreciate your cooperation in these i mer. ;

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I Edward Wenzinger, Chief f

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Reactor Projects Branch

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Enclosure:

Table of NU's Responses

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ENCLOSURE ., . WC.*

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,3 NRC NU RESPONSE SUBJECT OF ISSUE NUMBER RESPONSE DATE w4y 4 ,

91-070-1 A09805 09/27/91 NO PLANT INCIDENT REPORT FOR ~

'D' CIRCULATION PUMP TRIP i E 91-070-3B MOTOR HEATERS WERE NOT ENERGIZED ON SPARE BORIC ACID PUMP - .91-079 A09768 09/27/91 GAMMA METRICS PROCEDURES .

91-ll3N & A09699 09/13/91 ABNORMAL STACK RADIATION 91-136 MONITOR INDICATION 91-114-1 REPETITIVE FAILURES OF VENTE STACK HIGH RANGE MONITOR >

91-114-2 ACCUMULATOR TANK LEVEL CALIBRATION ERROR 91-116 ,

INADEQUATE RWP/HP CONTROLS 91 122 CLEAN LIQUID RADIATION ~

' ! '!ON! TOR TAGGING ERROR

__: 2 TCFSERV:CE 5 TACK ',10N;TCR '

IMPROPER ASSIGNMENT OF ON-CALL I&C TECHNICI AN 91-129 VOLUME COffrROL TANK LEVEL .j EVOLUTION l

'91-130 WRONG PART FOR LOCAL PRESSURE GAGE 91-162 A09698 09/09/91 UNCALIBRATED/ UNLABELED ,

EXHAUST VENTILATION DUCT, ,

MANOMETER l A09819' 09/27/91 WIDE RANGE NUCLEAR I 91-163 INSTRUMENT UNDOCUMENTED [

SWITCHES '

91 165-1 A09703 09/09/91 UNLABELED CRYOGENIC LIQUID 3 ~ l NITROGEN CONVERTER VALVES 6 l 91-165-2 PERMANENTLY CONNECTED . .

NOT DEPICTED PUMP 91-166 A09701 09/13/91 RECOVERY BORIC ACID TANK _ !

PUMP ISOLATED SWITCH i

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JAN 131392 j Docket No. 50-336 l Mr. John F. Opeka :

Executive Vice President - Nuclear ;

Northeast Nuclear Energy Company l P.O. Box 270 j Hanford, Connecticut 06141-0270 t

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Dear Mr. Opeka:

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Subject:

NRC Region I Inspection Repon No. 50-336/91-29

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Mr. J. T. Shediosky and others of this office conducted a special safety inspection' from .

October 21 through December 16,1991, at the Millstone Nuclear Station Unit 2, Waterford, Connecticut. The inspection results are documented in the enclosed repon. They were discussed with Mr. J. S. Keenan and other members of your staff at the conclusion of the- j inspection. i Areas examined during the inspection are described in the enclosed repon. Within these areas, j

( the inspection focused on issues brought to you by the NRC. Our independent review evaluated - j your performance in complying with regulatory requirements important to public and worker - l' health and safety. This review consisted of performance observations of ongoing activities.

intersiews with personnel, and review of records. ;

Our overall assessment was that NNECO's performance was acceptable. - Areas were identified w hich needed improvement. The enclosed inspection repon notes a number ofissues on which your staff agreed to provide a response to the NRC. Except where required for response to. I s iolations, the response to the NRC can be made in communications with the resident inspectors. ,

i A violation is discussed in the enclosed Notice which you are required to respond to and, in ~!

preparing your response, you should follow the instructions in the Notice. The violation mvolves multiple examples of individually minor problems with procedure compliance.

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.I Northeast Nuclear Energy Company 2 In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter and its enclosures will be placed in the NRC Public Document Room. The responses directed by this letter are not subjected to the clearance procedures of the Office of Maragement and Budget as required by the Paperwork Reduction Act of 1980, Public Law No. 96.511.

Your cooperatic, with us is appreciated.

Sincerely, Edward C. Wenzinger, ChIgj Projects Branch No. 4 Division of Reactor Projects

Enclosures:

1. Notice of Violation

2. NRC Region I Inspection Report No. 50-336/91-29 cc w/ enclosures:

W. D. Romberg, Vice President, Nuclear Operations D. O. Nordquist, Director of Quality Services R. M. Kacich. Manager, Nuclear Licensing S. E. Scace, Nuclear Station Director, Millstone J. S. Keenan, Nuclear Unit Director, Millstone Unit 2 Gerald Gar 6 eld, Esquire Nicholas Reynolds, Esquire K. Abraham. PAO (2)

Public Document Room (PDR)

Local Public Document Room (LPDR)

Nuclear Safety Information Center (NSIC)

NRC Resident Inspector State of Connecticut SLO Designee

3 A; .

4 U.S. NUCLEAR REGULATORY COhBilSSION REGION I 1; g, o ,

Report No.: 50-336/91-29 .

/ '

/ . .

DRP-65 License No.:

Licensee: Northeast Nuclear Energy Company P.O. Box 270 1 Hanford, CT 06141-0270 Facility: Millstone Nuclear Power Station, Unit 2 Location: Waterford, Connecticut j Inspection ,

Dates: October 21 - December 16,1991 Inspectors: T. R. Fredette, Consultant, AMSEC/SAIC .

T. G. Humphrey, Consultant, EG&G, INEL {

C. M. Meeker, Consultant, COMEX/SAIC - y P. L. Reagan, Consultant, COMEX/SAIC - !

(- E. L. Conner, Reactor Licensing / Risk Engineer, Technical Support Section, DRP Supervisor: J. T. Shediosky, Senior Allegation Coordinator, Reactor Projects Section No. 4A -

Division of Reactor Projects Approved by: L 61 fk.

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I Eugene Kelly, @i _

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cactor Projects S on 4A ,

Division of Reactor Projects

.s Scope: . Special inspection of concerns brought to the licensee by the NRC. This report is a continuation of the special inspection described in NRC Inspection Repon 50-245/91-23 and 50- '

336/91-27. It included the observations and evaluations during conduct.of surveillance and.

calibration activities and review for adequacy of maintenance procedures and procedure control .

issues. ,

Inspection Results: See Executive Summary ;

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SURVEILLANCE TESTING" section of this report contains detailed ,

conclusions regarding specific surveillance tests.

3.0 SURVEILLANCE TESTING Administrative Control Procedure ACP-QA-3.02E, section 6.2, stated that compliance is expected" for those procedures used tod doib surveillance d the Unit-2 Technical Specifications. The inspector observed surveillance testing as in the following.

3.1 Surveillance 2401B-1: Wide Range Nuclear Instrumentation Nuclear instrumentation at Millstone Unit 2 includes excore and Ten channels of excore instrumentation monitored neutron flux and provide :r Four of those channels are Gamma and control signals during startup and power operation.

Metrics wide range nuclear instrumentation (WRNI) designed to monitor rea Unit-2 Technical Specifications (TS) requires a '

source range to above 100 percent power.

minimum of two WRN1 channels to be operable in mode 6 during operations alteration or positive reactivity change and in modes 3, 4 and 5. l includes monitoring reactor power and removing the reactor protection system (RP tc

- mode bypass above 10' percent power.

The NRC provided a number of concems to the licensee related to WRNI The concerns related to intermittent spiking of the WRNI, I&C procedures for t NRC disposition of those concerns involved providing the concern to and related matters.

NNECO for review and resolution, with subsequent NRC evaluation to en NNECO's actions. Four NNECO letters (A09163, A09557, A09768, and A0 I NNECO's review of those concerns.

l 50-245/91-23 and 50-336/91-27 (IR 91-27), section 7.2,27described a l NRC Inspection Report number of concerns regarding the accuracy of as-built conditions shown in d.

concluded, in part, there were weaknesses in NNECO's coordination of ven controlled drawings.

f inspection of the WRNI at Unit 2 was to evaluate the l The purpose of this (50-336/91-29) adequacy of resolution of the purported spiking problem, review represe .

observe surveillance testing and inspect related issues.

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a. Intermittent Spiking T BackcrounJ Until July 1991, intermittent WRN1

spiking" occurred. Spiking was a condition involving ,

spurious indication by a WRNI channel. Inspection Repon 50-336/90-22, section 5.3.3, documented an NRC inspection of WRNI spiking that occurred during the 1990 refueling outage.

That report concluded NNECO met Technical Specification requirements and adequately monitored core conditions.

Assessment According to NNECO letter A09163, dated December 21,1990, NNECO implemented during 1989 and 1990 a total of six AWOs to troubleshoot and resolve the WRNI spiking problem. The Production Maintenance Management System (PMMS) listed AWOs M2-90-05376, M2 11146. and M2-90-11791 as examples of such efforts.

NNECO did a WRN1 operability evaluation December 12,1990. Although that evaluation was in response to environmental qualification concerns of detector cable assemblies, it discussed the requirements for operability which were applicable to a potentially degraded WRNI channel.

Plant Operations Review Committee (PORC) concurred with that operability evaluation, as documented in meeting minutes PORC 2-90-192.

A review of the PMMS data for 1990 and 1991 indicated there were no open work orders related to resolution of the WRN1 spiking problem. The last instance of completed work N regarding the WRNI spiking problem was documented in AWO-M2-91-06141, July 1,1991 and there were no WRN1 spiking problems reponed since that date. The problem was corrected by rebuilding a spare instrument drawer using the best available circuit cards. Since the equipment is obsolete, the licensee was tasked with selecting these from new warehouse spares, or from cards located in the spare or original drawer. Following maintenance, the spare drawer was calibrated and tested in accordance with SP 2401B and IC 24171.

The PMMS listed various work orders. unrelated to the spiking problem, planned for the WRNI system. For example, repair of the spare WRNI drawer and replacement of the Gamma Metrics ;

cable assemblies for WRNI channels B and D. Also, NNECO stated that because the original ;

sendor of the WRN1 drawers no longer supported that equipment, NNECO was considering a plant design change that would replace the WRNI drawers with new units in approximately 1993.

1 The inspector discussed WRN1 spiking with 1&C and Operations management personnel. They indicated the WRN1 spiking problem appeared to be resolved. Also, they stated the WRNI system was successfully used in the past five months during plant startup.

Conclusions Based on review of PMMS data, AWO-M2-91-06141 and the December 12,1990, WRN1 operability evaluation, discussion with cognizant NNECO personnel, and observation of WRNI L

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4 functional testing, the inspector concluded NNECO adequately resolved the WRN1 spiking f

/ problem.

1

b. Status of Procedures for WRNI I&C Work -!

The inspector reviewed the status of various procedures used to do I&C work on the WRNI. !

Assessment ,

The inspector identified and did a general review of one surveillance procedure (SP) and five maintenance (IC) procedures that applied specifically to the WRNI. They were SP 2401B, IC 2416G, IC 2417C, IC 2417D, IC 2417H, and IC 24171. NNECO revised two of these l procedures (SP 2401B and IC 2417H) based on the procedure upgrade program, while three [

were in the upgrade process, and IC 24171 was not upgraded. NNECO stated its intent was to j upgrade all such PORC approved procedures by the end of 1992.

The biennial review of IC 2416G was overdue. ACP-QA-3.02D, section 6.1.1, required a .

periodic, systematic review of Station Procedures required by ACP-QA-3.02. ACP-QA-3.02, !

e section 6.2.3, included 2400 series SP or IC procedures. In a quarterly memorandum (MP [

918), dated November 1,1991, Document Services identified the last biennial review date for {

IC 2416G and IC 2417C as September 1,1987, and August 1,1989, respectively. I&C l , management stated I&C Department records indicated the last date for the biennial review of -

~

those procedures was September 1989 and December 1990, respectively.

l

. i ACP-QA-3.02, section 6.8.1, noted a general rule that "after three changes have been made to ;

any procedure, a revision should be written to incorporate those changes." Procedure IC 24171 ,

had four changes and form IC 24171-1 had six changes. The I&C Department was aware of the need to revise IC 24171. The I&C Department priority order for procedure revision was the -

following: " problem" procedures surveillance procedures, radiation monitor procedures, procedures with more than three changes, and numerical order.

L i

Conclusions

! Based on review of applicable procedures, the inspector concluded that procedures for WRNI l

!&C work were adequate and that NNECO was in the process of improving those procedures. {

All biennial reviews were not completed as required by ACP-QA-3.02D. This was a violation ;

of NRC requirements (VIO-50-336/91-29-01). i Also, because some Document Services and 1&C Department records of biennial reviews l differed, the inspector questioned if NNECO tracked biennial reviews in accordance with ACP- ;

QA-3.02D, section 6.3. NNECO agreed to evaluate this matter, take appropriate action as necessary, and respond to the NRC.

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c. Conduct of WRN1 Surveillance 2401B-1 '

November 13,1991, the inspector observed the conduct of Unit-2 surveillance 2401B-1 that was done under AWO M2-91-12287. Since the reactor was in mode 5, TS required performance of h

SP 2401B. The purpose of routine weekly surveillance 24018-1 was to verify operability of all four WRN1 channels.

i Assessment With the following two exceptions noted for WRNI channel A, the I&C Technician successfully completed surveillance 2401B-1 in a thorough and diligent manner. First, step 6.1.17.2 required removal of fuses F1 and F2 on the power supply mounting plate. Only fuse F1 was removed. !

The 1&C Technician doing the surveillance stated that, in this case, removal of only fuse F1 was adequate to achieve the expected resuh (de-energization of the wide range detector high voltage power supply).

i The inspector noted that drawing J178-0010, Interconn Schematic Wide Range Channel h1W3, revision k, indicated fuses F1 and F2 were for the power feeds to ungrounded power supplies ,

PSI and PS2. Therefore, the procedural requirement may have been based on personnel safety in dealing with an ungrounded power supply.

Second, step 6.1.19 was done before step 6.1.18 and the procedure did not allow steps to be ...

done out of sequence. The inspector discussed the above exceptions with the I&C Technician >

and the I&C Technician agreed the inspector's observations were factually correct.

3 Subsequently, surveillance of WRNI channels B, C, and D was done in accordance with SP 2401B and 1&C Form 2401B-1.

1 The inspector discussed the preceding exceptions with I&C and Quality Senices Division (QSD) i management. QSD stated that it noted similar exceptions in other procedures during the on-going procedure compliance program.

I ,

An additional issue concernmg surveillance 2401B-1 was the completeness with which the procedure required checking control room annunciator responses during the suneillance. The procedure appropnately checked panel C04 annunciator window A12B, NIS Channel INOP, '

during step 6.1.8.1. The procedure did not check that panel C04 annunciator window Al2B cleared at step 6.1.10. There were other steps (e.g., 6.1.17.1 and 6.1.17.6) that actuated or cleared panel C04 annunciator window A12B, but the procedure did not require an annunciator I status check for each actuation and reset condition. Also, section 6.1 of the procedure did not i require checking all affected annunciators (e.g., panel C04 annunciator window C12A, CH 'A' Wide Range Extended Range C.P.S.).

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Conclusions '

Based on observation of the conduct of surveillance testing done using SP 2401B and resiew of completed 1&C Forms 2401B-1, the inspector concluded that NNECO adequately completed surs eillance 24018-1 on November 13, 1991. The two noted exceptions had no significant

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regarding procedural compliance as required by ACP-QA-3.02E. The failure to follow !

procedure SP 2401B is a violation of NRC requirements (VIO-50-336/91-29-01).

The completeness of annunciator response checks in SP 2401 and I&C Form 2401B-1 may not -

be adequate. NNECO agreed to evaluate this matter, take appropriate action as mary, and ,

respond to the NRC. ;

f The !&C Technician demonstrated a high level of skill and knowledge during the conduct of this

[

surveillance. The inspector considered this to be a strength. i

d. Qualification of I&C Technicians :

To sample the OJT qualification status process, the inspector requested NNECO to produce for I inspection the qualification records for the I&C Technician who did SP 2401B-1 November 13, !

1991. Records existed to confirm this' person was the only I&C Technician who had !

documentation of qualification to do this surveillance. :

t Due to illness, the SP 2401B-1 qualified individual was not at MP2 December 3,1991, when SP 2401B-1 was done by another I&C Technician. At approximately 1314 hours0.0152 days 0.365 hours 0.00217 weeks 4.99977e-4 months , while in the' l n MP2 Control Room on an unrelated inspection, the inspector briefly observed performance of .

SP 2401B-1 under AWO M2-91-12853. The inspector questioned I&C management if the. !

second I&C Technician had documentation of qualification to do SP 24018-1. I&C management i 4

stated the I&C Technician, as shown by the qualification matrix, did not have documentation of ;

qualification to do SP 2401B-1. A Supervising Control Operator log book entry at 1400 hours0.0162 days 0.389 hours 0.00231 weeks 5.327e-4 months indicated completion of SP 2401B-1. After completion of SP 2401B-1, I&C management ,

initiated an I&C Form 2450-6, Justification For Use of Individual Not Having Documented. !

Qualification, for the person who did AWO M2-91-12853. . l The inspector reviewed the I&C training folder for the I&C Technician who did AWO M2 12853 and discussed the I&C Technician qualification process with the Nuclear Training - 1 "

Depanment (NTD). According to NTD records, the I&C Technician who did AWO M2 12853 had appropriate qualifications to do that work. The inspector also. reviewed i

documentation of qualification status for other I&C Technicians, as subsequently described in :

this report section. '

The inspector noted that department procedure IC 2450, section 6.6.2.3, required an annual !

proficiency review in accordance with ACP-QA-8.16 for level.II certified I&C personnel. j However, ACP-QA-8.27 and ACP-QA-8.29 superseded ACP-QA-8.16 on February 1,1991. :

Procedure IC 2450 has not yet been revised to delete the requirement for proficiency reviews. !

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, Conclusions Based on inspection of I&C Department and NTD qualification records for several individuals '

and review of applicable procedures, the inspector concluded the I&C Department did not x' >

adequately maintain and use the qualification matrix in all cases to determine the qualification status of I&C Technicians. Also, administrative requirements for I&C personnel certification in IC 2450 were not entirely consistent with the curTently applicable ACPs. NNECO agreed to evaluate this matter, take appropriate action as necessary, and respond to the NRC.

3.2 Surveillance 2401F-1: Reactor Protection System High Power Trip The Reactor Protection System high power trip (RPS-HPT) test is a suneillance to ensure operability of the variable high power trip calculator, the nuclear power - delta T power max select unit, and their associated functions.

a. Conduct of Surveillance Testing The inspector observed the conduct of surveillance 2401F-1, for RPS-HPT channels A and B, done December 4,1991, under AWO M2-91-ll846. This surveillance was done by two I&C Technicians using a reader-doer method. t Assessment h MP2 TS 3.3.1.1 requires a minimum of three operable RPS-HPT channels in modes 1,2, and 3, except when all control rod drive mechanisms are de-energized or when the RCS boron ;

concentration exceeds the specified refueling concentration. To demonstrate operability of the :

RPS-HPT when in modes 1,2, or 3 (with reactor trip breakers closed), MP2 TS Table 4.3-1 l includes requirements for a channel check once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and a channel calibration and a !

channel functional test once per 31 days. The channel calibration and the channel functional test were done using SP 2401F and Form 2401F-1. MP2 was in mode 5 and this surveillance was l done as a pre-start check, since plant startup was anticipated within one week. l The inspector observed the conduct of the surveillance test. With te following five exceptions regarding procedure steps either being done out of sequence or not being done exactly as specified, I&C Technicians successfully completed surveillance 2401F-1 for RPS-Hirr channels A and B in a thorough and diligent manner. First, *as left" data for step 6.1.3 was mistakenly recorded in the wrong location on the data sheet for Channel A. This data was appropriately recorded later. Second, steps 6.2.1 through 6.2.4 for Channel B were not done in the specified sequence. Third, for Channel A, recording of data at step 6.4.11 was inadvertently omitted.

Later, the step was repeated and data was recorded. Fourth, for Channel B, the high power level trip bistable was not reset in step 6.4.11. Fifth, for Channel B, the CPC #2 test probe was restored to its storage position at step 6.4.2 rather than at step 6.10.7. t 6

$l RECORD OF ALLEGATION PANEL DECISIONS SITE: Mr <; %4 PANEL ATTENDEES:

i ALLEGATION NO.: 21- 9 d- A- 00 82 Chairman - ~J 'T u>cu s M !

DATE: 944av93 (Mtg. I 2 3 4 5) Branch Chief - E. {_ UJo oNa pe. :

PRIORITY: High Low Section Chief (AOC) - E.M.)/efh SAFETY SIGNIFICANCE: Yes No Unknown Others - 2. C. L-CONCURRENCE TO CLOSE00T: DD @ SC 2.6r.Naa CONFIDENTIALITY GRANTED: Yes No k.k. vmeif ce (See Allegation Receipt Report)

IS THEIR A DOL FINDING: Yes No IS CHILLING EFFECT LETTER WARRANTED: Yes No '

HAS CHILLING EFFECT LETTER BEEN SENT: Yes No HAS LICENSEE RESPONDED TO CHILLING EFFECT LETTER: Yes No ACTION: .

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Docket Nos. 50-336 File Number RI-91-A-0082 i

Northeast Nuclear Energy Company NITN: Mr. E. J. Mroczka Senior Vice President - Nuclear Engineering and Operations Group P.O. Box 270 -

Hartford, Connecticut 06141-0270

Dear Mr. Mroczka:

The U.S. Nuclear Regulatory Commission recently received information concerning activities at the Millstone Nuclear Power Facility, Unit 2. The details are enclosed for your review and follow-up.

We request that the results of your review and disposition of these matters be submitted to Region I within 30 days of the date of receipt of this letter. We request that your response contain no personal privacy, proprietary, or safeguards information so it can be released to the public and placed in the NRC Public Document Room. If necessary, such information shall be contained in a separate attachment which will be withheld from public disclosure.

The affidavit required by 10 CFR 2.790(b) must accompany your response if proprietary information is included. Please refer to file number RI-91-A-0082 when providing your response.

The enclosure to this letter should be controlled and distribution limited to personnel with a "need to know" until your investigation of the concern has been completed and reviewed by NRC Region I. The enclosure to this letter is considered Exempt from Public Disclosure in accordance with Title 10 Code of Federal Regulations, Part 2.790(a). However, a copy of this letter excluding the enciasure will be placed in the NRC Public Document room.

The response requested by this letter and the accompanying enclosure are not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511.

Your cooperation with us is appreciated. We will gladly discuss any questions you have concerning this information.

Sine ely, .

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C . les W. , D ector Division of Reactor Projects

Enclosure:

(10 CFR 2.790(a) Information) r . -

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~ LIMITED,DISTRIBUTIONjNOT FOR'PUBI.lC-DISC 10SURTu-Issue 1:

ne wiring diagrams invohing Reactor Coolant Pump RTD circuits have not been updatedfollouing modifcations made under a PDCR to replace RTD circuit kmfe '

switches with Weidmuller Test Blocks. Drawing No. 25203-31069, Sheet 5, Rev. 3, dated August 29,1989, does not reflect the changefor at least 4 RTD circuits (TCD, TCC, TCA, TCB). Ihe instrument loop diagrams (Drawing No. 25203-28500, Sheets ,

140 & 146) show the Weidmuller Test Blocks. Also, in Drawing number 25203-31069, Sheet 5, thejumpers shown between cable lead 1 and the cable shield ground on the loop diagrams are not shown. In addition, access to the GRITS systen, to venfy the latest drauing revisions, is restricted in that personal access codes are only validfor 30 days.

Request 1:

Please discuss the validity of the above assertions. If discrepancies are found, please assess the significance of the discrepancies with respect to plant operation and safety and discuss any actions taken or planned to correct these discrepancies.

Issue 2: ,

he Steam Generator No. 2 mid-loop instntmentation (L-122) was not " operable" during drain-downfor tube inspections on May 2,1991. GEM switches uerefound to be ' frozen' on in place. In addition, L-112 had an electronic noise problem caused ,

by an improperly installedjumper. 7hus licensee commitment that two monitors be operable during drain down condition uns 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 the Steam Generator No. 2 drain-down evolution. Also please discuss any actions taken or planned to correct these deficiencies.

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i Pressure indicating instnanent (PI 6350 A/B and PI 6351 A/B) and mountingsfor {

ser ice unter (SW) supply to emergency diesel generators (EDG) are not seismically l mounted. Any kind ofshock uvuld be suficient to knock the gauge and valve of of ;

the strainer. Additionally, the location of the taps as shown on the P&lD apparently does not coincide uith the actual tap locations. l Request 3:

Please discuss the validity of the above assertions. If the assertions are valid, please ,

discuss their effect on the safe operation of SW supply to the EDG. Please provide i any actions taken or planned to ensure that seismic requirements for these instruments ;

are being met.

Issue 4:

On Afay 3,1991, the Unit 2 Stack Radiation Afonitor (Rhi 8132) uns inoperable as a .l result of beingfooded with water. This monitor uvuld have been inoperable anyuny, i as airflow had been isolated. Filling and pressure testing ofSteam Generator (SG) {

1 was underuny during the same time period. Problems with valve line-upsfor the rad monitor and the SG testing contributed to the flooding and monitor inoperability.

Additionally, health physics (HP) controls during removal of the waterfrom the monitor uns inadequate resulting in contamination ofpersonnel.

Request 4:

-l Please discuss the validity of the above assertions. If discrepancies are confirmed, please discuss actions that you have taken or will take to ensure that plant procedures '

regarding rad monitor operation, conduct of tests, and HP activities are being used properly.

Issue 5:

Procedure discrepancies exist beturen OP-2336E and SP-2617A for the restoration of the line-up of the radiation monitor (RE-245), and its associated sample pump. t u

Operators routinelyfail to perform OP-2336E, Section 5.1, Step 5.1.13 which is to immediately close AOV-244A/B and AOV-245 when securingfrom condensate polishingfacility discharges. Thisfailure tofollowprocedures results in the sample pump to radiation monitor (Rhi-245) continuing to operate when the tank discharge is secured.

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Request 5:

Please discuss the validity of the above assertions. If any discrepancies are identified concerning procedure noncompliance, please discuss their significance on the operation of radiation monitor RM-245. Please discuss any corrective actions taken or planned, to ensure operators are meeting procedural and technical specification requirements.

Issue 6:

Thefollowing discrepancies have been idennfied during an evaluation of Work Order AWO-Af2-91-M411. These discrepancies idennfy continued non-compliance with ;

procedures and poor response of operations and management to recurring problems with radiation monitor RE-245.

a. The sample pump continues to run when the tank discharge stops at 15% tank ,

level (TK-11). l

b. The

Low Flow" switch does not always see a lowflow condition when TK-10 and TK-11 discharge pumps stop. The head of water in the pipe and tidal conditions afect theflow of water.

c. Operations normally rely on the 15% tank level pump trip to stopflow causing a lowflow to trip shut RE-245 discharge valve, and AOV-245. If AOV-244A/B

' are shut and no lowflow condition exists, RE-245 sample pump will condnue to run until AOV-245 is shut.

d. Changes to OP-2336E wre idennfied in 1989 to prevent the problems idennfied by AWO-Af2-91-W411. However continued idennfied procedure non-compliance 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 of radiation monitor RE-245. Please discuss any corrective actions that are being used to correct the problems. ,

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JUN 0 71991 Docket Nos. 50-336 File Number RI-91-A-0082 Northeast Nuclear Energy Company ATTN: Mr. E. J. Mroczka Senior Vice President - Nuclear Engineering and Operations Group P.O. Box 270 Hartford, Connecticut 06141-0270

Dear Mr. Mroczka:

The U.S. Nuclear Regulatory Commission recently received information concerning activities at the Millstone Nuclear Power Facility, Unit 2. The details are enclosed for your review and follow-up.

We request that the results of your review and disposition of these matters be submitted to Region I within 30 days of the date of receipt of this letter. We request that your response contain no personal privacy, proprietary, or safeguards information so it can be released to the public and placed in the NRC Public Document Room. If necessary, such information shall be contained in a separate attachment which will be withheld from public disclosure.

The affidavit required by 10 CFR 2.790(b) must accompany your response if proprietary information is included. Please refer to file number RI-91-A-0082 when providing your response.

The enclosure to this letter should be controlled and distribution limited to personnel with a "need to know" until your investigation of the concern has been completed and reviewed by NRC Region 1. The enclosure to this letter is considered Exempt from Public Disclosure in ,

accordance with Title 10, Code of Federal Regulations, Part 2.790(a). However, a copy of this letter excluding the enclosure will be placed in the NRC Public Document room.

The response requested by this letter and the accompanying enclosure are not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511. :

i i

Your cooperation with us is appreciated. We will gladly discuss any questions you have concerning this information.

Sincerely:, ad 3y p;u i

% LL)q qsM Charles W. F ehl, Di .' tor Division of Reactor Projects

Enclosure:

(10 CFR 2.790(a) Information) ,

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Letter to NNECO on Allegation RI-91-A-0082 i

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Imcal Public Document Room (LPDR)

State of Connecticut t i

bec w/ encl: 5 Allegation File,(6) RI-91-A-0082,0084,0085,0096,0099,0H6 T. Shedlosky, SRI, Millstone J. Stewart E. Kelly to . R p..A f

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AUG 0 61331 Docket Nos. 50-336 File Number RI-91-A-0079 Northeast Nuclear Energy Company ATTN: Mr. E. J. Mroczka Senior Vice President - Nuclear Engineering and Operations Group P.O. Box 270 Hartford, Connecticut 06141-0270 ,

1

Dear Mr. Mroczka:

The U.S. Nuclear Regulatory Commission recently received information concerning activities '

at the Millstone Nuclear Power Facility, Unit 2. The details are enclosed for your review and followup. .

We request that the results of your review and disposition of these matters be submitted to Region I within 30 days of the date of receipt of this letter. We request that your response contain no personal privacy, proprietary, or safeguards information so it can be released to :

i the public and placed in the NRC Public Document Room. If necessary, such information shall be contained in a separate attachment which will be withheld from public disclosure. !'

The affidavit required by 10 CFR 2.790(b) must accompany your response if proprietary information is included. Please refer to file number RI-91-A-0079 when providing your l response.

The enclosure to this letter should be controlled and distribution limited to personnel with a

need to know" until your investigation of the mocern has been completed and reviewed by NRC Region I. The enclosure to this letter is considered Exempt from Public Disclosure in '

accordance with Title 10. Code of Federal Regulations, Part 2.790(a). However, a copy of this letter excluding the enclosure will be placed in the NRC Public Document room.

The response requested by this letter and the accompanying enclosure are not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511. ,

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Your cooperation with us is appreciated. We will gladly discuss any questions you have .

concerning this information.

Sincerely, Ch W. Hehl, D ector :

Division of Reactor Projects ;

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Enclosure:

(10 CFR 2.790(a) Information) cc w/o encl:

Public Document Room (PDR) l Local Public Document Room (LPDR) .

State of Connecticut i

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Issue 79-1 '

Procedures being issued in the I&C department are inadequate in that acceptance criteria are not being established for required measurements. Specifically a draft copy of procedure IC 2416G was provided for review and it had not incorporated several comments that were raised on previous revisions. These comments included: 1) An acceptance criterion for the !

output of the pulse height discriminator was not established; 2) A precaution was not added to check the power supply output of the NLW-3 drawer if the Gammametrics power supply drops below 15 volts; The Gammametrics and NLW-3 drawers share the same power supply and the Gammametrics output acceptance criterion is 15 i 1.5 VDC while the NLW-3 output acceptance criterion is 15 i 0.0075 VDC, Therefore the Gammametrics drawer may be in ;

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specifica': ion while the NLW-3 is out of specification; 3) The proposed acceptance criterion for the discriminator bias voltage was inadequate at .9 i 1 VDC; Gammametrics recommends 0.8 to 1.0 VDC.

In addition PORC meeting 2-89-123 authorized change No. 3 to procedure IC-2417I-1. The i

change tuthorized new settings for NLW-3 drawer discriminator voltage. Section 5.4 of IC-24171-1 should have also been changed at this time, and a discussion section on NLW-3 discriminator settings should have been added.

Request 79-1 Please discuss the validity of the above assertions. Was the procedure released for use, and if so, was it unusable in this field? Please state whether or not the procedural changes were ,

required. to satisfy regulatory requirements, and discuss the review process for procedures and

how comments raised during the procedure review are addressed.

t Issue 79-2 While troubleshooting a disabled ICC thermocouple, it was noted that a Litton-Veam connector used to perform the troubleshooting was identical to what is installed in the ICC system. The connector used for the tmubleshooting was obtained from the NNECO warehouse and was not EQ. Work order AWO M2-90-13287, used for the troubleshooting, did not reference procedures IC 2421C and IC 2821E, which provide guidance to personnel for working on Litton-Veam connectors that are EQ. I&C department supervisors were unaware that this section of the ICC system cabling was EQ. Also the loop folders which were being used for this work were out of date.

Request 79-2 Please discuss the validity of the above assertions. If valid, please discuss actions taken to ensure that EQ requirements were met in this case.

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Issue 79-3 l Recently, a PDCR which installed an audio monitoring system on the pressurizer safety i valves was authorized. The audio system did not contain a spare hookup as shown on the l PDCR drawing. Also the wire hookup in the PDCR showed two different setups. These problems caused the job to be delayed resulting in excessive radiation exposure of the :

workers. ;

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Request 79-3 i Please discuss the validity of the above assertions. If valid, please discuss the methods used !

to ensure that procedures are technically correct prior to performance and what preparations are performed to ensure worker radiation exposure is minimized.

Issue 79-4 On April 22,1991 the "B" and "D" battery chargers were removed from their normal power ;

i supply.12ter, a technician performing a surveillance on the "C" train Nuclear Instrument i placed the instrument into Test due to a spiking problem. This action rendered three trains [

of nuclear instrumentation inoperable. Since the plant was in Hot Standby at the time of the i' test, two trains of nuclear instrumentation were required to be operable. A Plant incident Repon (PIR) was initiated to document the cccurrence but the PIR was later canceled based :

upon an interpretation of the Technical Specification requirements. j i

i Request 79-4 Please discuss the validity of the above assertions. Please discuss the basis for the decision ;

not to document the occurrence, if substantiated, with a PIR. (

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4 Docket Nos. 50-336 File Number RI-91-A-0079 1

Northeast Nuclear Energy Company ATTN: Mr. E. J. Mroczka Senior Vice President - Nuclear Engineering and Operations Group P.O. Box 270 Hartford, Connecticut 06141-0270

Dear Mr. Mroczka:

The U.S. Nuclear Regulatory Commission recently received information concerning activities i

at the Millstone Nuclear Power Facility, Unit 2. The details are enclosed for your review and followup.

We request that the results of your review and disposition of these matters be submitted to Region I within 30 days of the date of receipt of this letter. We request that your response contain no personal privacy, proprietary, or safeguards information so it can be released to the public and placed in the NRC Public Document Room if necessary, such information shall be contained in a separate attachment which will be withheld from public disclosure. ,

The affidavit required by 10 CFR 2.790(b) must accompany your response if proprietary information is included. Please refer to file number RI-91-A-0079 when providing your response.

The enclosure to this letter should be controlled and distribution limited to personnel with a "need to know" until your investigation of the concern has been completed and reviewed by NRC Region 1. The enclosure to this letter is considered Exempt from Public Disclosure in accordance with Title 10, Code of Federal Regulations, Part 2.790(a). However, a copy of this letter excluding the enclosure will be placed in the NRC Public Document room.

The response requested by this letter and the accompanying enclosure are not subject to the clearance procedures of the Of6ce of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511.

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N: NORTHEAST UTILITIES cene,,, On,ce, . see n stro,,. B.,,,n. Conn cucm m um uw w u.o .u n 1 .nw ~c~w"5 w*"** P O. BOX 270 N.,b,7,Yeh, HART FORD. CONNECTICUT 06141-0270 k ' J w .c u-ew c-m (203) 665-5000 August 9, 1991 Docket No. 50-336 A09604 Mr. Charles V. Hehl, Director Division of Reactor Projects U.S. Nuclear Regulatory Commission '

Region I 475 Allendale Road King of Prussia, Pennsylvania 19406 ,

Dear Mr. Hehl:

Millstone Nuclear Pover Station, Unit No. 2 RI-91-A-0082 Ve have completed our review of the identified issues concerning activities j at Millstone 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. l The NRC letter and our response have received controlled and limited distribution on a "need to know" basis during the preparation of this ;

response.

ISSUE 1: -i The viring diagrams involving Reactor Coolant Pump RTD circuits have not been updated following modifications made under a PDCR to replace RTD ,

circuit knife switches with Veidmuller Test Blocks. Draving No.

25203-31069, Sheet 5 Revision 3, dated August 29, 1989, does not reflect the change for at least 4 RTD circuits (TCD, TCC, TCA, TCB). The instrument loop diagrams (Drawing No. 25203-28500, Sheets 140 & 146) shov l 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 not shown. 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.

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.k Mr. Charles V. Hehl, Director U. S. Nuclear Regulatory Commission A09604/Page 2 August 9, 1991 Request 1:

Please discuss the validity of the above assertions. If discrepancies 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.

For clarity, our response to this issue is segregated into two parts. Part A 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. knife switches with Veidmuller Inc. Test Blocks. As a result of the PDCR, 320 instrument loops were modified which required 330 drawing changes. Draving 25203-31069 Sheet 5 was not changed at that time and, therefore, was not updated. Drawing 25203-39045, Sheet 55B includes all of the information of Drawing 25203-31069 Sheet 5 plus internal cabinet viring. Drawing 25203-39045 Sheet SSB vas updated at the time of the change and therefore does shov Veidmuller Test Blocks.

Response

The assertion that Drawing 25203-31069 Sheet 5 was not upgraded at the time of the PDCR implementation is valid. This was the result of an isclated oversight and is not indicative of a program deficiency. Drawing 25203-31069 Sheet 5 is being changed to show the Veidmuller Test Blocks and jumper configuration under Drawing Change Request DCR M2-S-1216-89.

PART B

Background:

Each individual with a need to access the GRITS system has been assigned a User Identification number by the Information 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 how 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. If system difficulties are encountered, an IRG HELP phone line is available (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months a day) as is department assistance.

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Mr. Charles V. Hehl, Director ;

U. S. Nuclear Regulatory Commission !

A09604/Page 3 i August 9, 1991 l t

Response

i The user ID and password system is designed to provide the necessary level f of security along with an appropriate level of ease of use for the person !

using the system. Adequate support for infrequent users is also provided. l l

ISSUE 2: j 1

The Steam Generator No. 2 mid-loop instrumentation (L-122) vas not ;

" operable" during drain-down for tube inspections on May 2, 1991. GEM switches were found to be " frozen" on in place. In addition, L-112 had an j electronic noise problem caused by an improperly installed jumper. Thus '

licensee commitment that two monitors be operable during drain-down i condition was not being met.

1 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-down ;

evolution. Also please discuss any actions taken or planned to correct i these deficiencies.

l Response: +

For additional clarity, this issue has been segregated into four sections. l A. The Steam Generator No. 2 mid-loop instrumentation (L-122) was not j

" operable" during drain-down for tube inspections on May 2, 1991.

B. GEM switches were found to be " frozen" in place, i

C. L-112 had an electronic noise problem caused by an improperly installed jumper. l D. The licensee commitment that two monitors be operable during drain-down i condition was not being met. i

Background:

A. During the. April /May 1991 Steam Generator shutdown, the Vestinghouse Ultrasonic level measuring system (L-122) vas not operable during drain-down for tube inspection. L-122 was procedurally deleted from use on April 24, 1991 (Procedure Change #3 to Operations Procedure OP 2301E, Rev. 15). Drring the April /May 1991 shutdown, Westinghouse provided, installed, and tested enhanced design transducers and softvare. Testing of the system continues until reliable system.

operation is achieved.

B. The GEM level indicator (LG-112) uses a floating magnet to position

" flags" that provide a visual indication of hot leg level. These flags are monitored by closed circuit TV in the Control Room. During the

a

? Mr. Charles V. Hehl, Director U. S. Nuclear Regulatory Commission A09604/Page 4 August 9, 1991 unplanned April /May 1991 outage, previously identified problems with .

the proper response of the flags to the magnet vere investigated. !

These problems were attributed to factory mismarked replacement flag assemblies and were corrected. The assembly was then tested and demonstrated as acceptable performance.

Subsequent to this activity, during mid-loop conditions during the April /May 1991 outage, the response of the GEM level indicator was observed to not change during very small (<3/4"), and slow changes in RCS level 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 was not observed during post installation testing performed during the 1990 !

refueling 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 standpipe that contains an internal guide rod with a new design that eliminates the ,

potential for guide rod binding. NNECO intends to obtain and replace the existing assembly with the new design in the future.

t C. LT-112 is a level sensor that generates an analog signal representing I 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 outage, unacceptable performance was noted and this feature vas defeated by installing a jumper. 7 During the April /May 1991 steam generator shutdown, unexplained bias and lov frequency output indication variations vere observed. Upon further testing and consultation with the manufacturer it was determined that the location of the jumper did not completely eliminate the interference of the lead circuit. The jumper placement was corrected and the system response stabilized.

The bias errors observed during the January unplanned outage vere attributed to inadequately-sized head vent tubing, and vere observed 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 evaluation for appropriate corrective action.

Response

A. L-122 was procedurally deleted from use on April 24, 1991, and therefore was not required for drain-down during the April /May 1991 shutdown. Troubleshooting efforts vere continued in a priority basis i e

'e 5 Hr. Charles W. Behl, Director 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 GEMS sensor was not found " frozen" in place as asserted. Poor response 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 jumper 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-down activities. At all times while draining to reduced inventory conditions, at least two level indication systems vere in operation. These systems satisfied the level monitoring requirements that were in effect.

ISSUE 3:

Pressure indicating instrument (PI 6350 A/B and PI 6351 A/B) and mountings for Service Water (SV) supply to Emergency Diesel Generators (EDG) are not seismically mounted. Any kind of shock would be sufficient to knock the gauge and valve off the strainer. Additionally, the location of the taps as shovn on the P&ID apparently does not coincide with the actual tap locations.

Request 3:

Please discuss the validity of the above assertions. If the assertions are valid, please discuss their effect on the safe operation of SV supply to the EDG. 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 drawing accuracy was previously identified to management. The design was reviewed and found acceptable for both dead weight and seismic loads. The drawing was 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 8132) was inoperable as a result of being flooded with vater. This monitor vould have been inoperable anyway, as air flow had been isolated. Filling and pressure testing of Steam Generator (SG) #1 was underway during the same time

i i Hr. Charles V. Behl, 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 you have taken or vill take to !

ensure that plant procedures regarding radiation monitor operation, conduct 7 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 water coming out of the vent upstream of the No. 1 Atmospheric Dump Valve, and flowing onto RM-8132.

The vent was closed to stop the water flow. 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 week, 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 was intended to be reviewed to indicate the actual desired position of the valve during operations. Procedure Writers Group individuals that were t involved thought that a change would be put into the valve line-up by the operating shift and that the new revision vould follow on a r.ormal schedule. The change was not submitted, and the next test was completed with the valve in the incorrect position which allowed water to flov'on to RM-8132, causir,g its failure.

The valve line-up errors vere 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 ;

associated sample pump. Operators routinely fail to perform OP 2336E, ;

Section 5.1, Step 5.1.13 vhich is to immediately close A0V-244A/B and ;

A0V-245 when securing from Condensate Polishing Facility discharges. This failure to follow procedures results in the sample pump to radiation monitor (RM-245) continuing to operate when the tank discharge is secured.

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9 5 Mr. Charles V. Hehl, Director U. S. Nuclear Regulatory Commission A09604/Page 7

. August 9, 1991 Request 5:

Please discuss the validity of the above assertions. If any discrepancies are identified concerning procedure noncompliance, please discuss their significance on the operation of radiation monitor RM-245. Please discuss any corrective actions taken or planned, to ensure operators are meeting procedural and Technical Specification requirements.

Response

See below.

ISSUE 6:

The following discrepancies have been identified during an evaluation of Vork Order AV0-M2-91-04411. These discrepancies identify 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 when the tank discharge stops at 15%

tank level (TK-11).

B. The " Low Flow" switch does not always see a flow condition when TK-10 -

and TK-11 discharge pumps stop. The head of water in the pipe and tidal conditions affect the flow of water, t

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 2330L were identified in 1989 to prevent the problems ;

identified by AVO-M2-91-04411. 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. ;

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9 i Mr. Charles V. Hehl, Director U. S. Nuclear Regulatory Commission i

A09604/Page 8 August 9, 1991 ,

i After our review and evaluation of the completed issues (Issues 1 through j 4), we find that these issues did not present an indication of a compromise ;

to nuclear safety. A valve line-up error was clearly made and it has been f corrected. Ve recognize the need to strive for a higher level of ;

performance in this area and we are aggressively working towards this obj ect ive. Ve appreciate the opportunity to respond and explain the basis j for our actions. Please contact my staff if there are any further questions on any of these matters. i l

Very truly yours, 1

NORTHEAST NUCLEAR ENERGY COMPANY

// /

E. Jj,f/ r6c'zk'a '

SenMr Vice President

/

ec: V. J. Raymond, Senior Resident Inspector, Millstone Unit Nos. 1, 2, and 3 E. C. Venzinger, Chief, Projects Branch No. 4, Division of Reactor Projects E. M. Kelly, Chief, Reactar Projects Section 4A l

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Dear Mr. Hehl:

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Dear Mr. Behl:

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Dear Mr. Venzinger:

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Dear Mr. Mroczka:

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Dear Mr. Behl:

Dear Mr. Opeka:

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Dear Mr. Opeka:

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Dear Mr. Mroczka:

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Dear Mr. Mroczka:

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Dear Mr. Mroczka:

Dear Mr. Hehl:

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