IR 05000245/1990007

Insp Rept 50-245/90-07 on 900403-0514.No Violations Noted. Major Areas Inspected:Plant Operations,Radiological Controls,Maint/Surveillance,Security,Engineering/Technical Support & Safety Assessment/Quality Verification

ML20055G185
Person / Time
Site:	Millstone
Issue date:	07/06/1990
From:	Haverkamp D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20055G183	List: IR 05000245/1990007 ML20055G182
References
50-245-90-07, 50-245-90-7, NUDOCS 9007200219
Download: ML20055G185 (22)

Text

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V.S. NUCLEAR REGULATORY COMMISSION

REGION I

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Report No~. 50-245/90-07-Docket No. 50-245

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"I License No. DPR-21'

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- Licensee: Northeast-Nuclear Energy. Company

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P.O.-Box 270

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Hartford, Connecticut '.06141-0270

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. Facility.Name: Millstone Nuclear Power-Station, Unit 1 Inspection'At:

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Waterford,. Connecticut-o.

iDates:' April 3 through May 14, 1990 Reporting Inspector:

D. Dempsey, Resident Inspector, Millstone _1

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Inspectors: W.J.lRaymond,, Senior Resident Inspector; 10; A Dempsey, Resident Inspector, Millstone l-

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1 Approved: By: _

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-Denald_R.- Haverkamp, Chie Date

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Reactor Projects Section A-Divi: ion.of Reactor Projects w

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Inspection Summary:

! Areas = Inspected: -Routine NRC. resident inspection of plant operations,

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radiological: controls,' maintenance / surveillance, security, engineering /

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- technical support, and safety assessment / quality verification.

!Results:' See Executive Summary

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EXECUTIVE SUMMARY

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L-s Plant Operations-i Routine review of this area identified strengths regarding plant configuration control 7and outage planning during the service water / condensate systems outage

,yt en_ April-2-7, 1990 (Section.3.4),

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'Radiologica'l Control s.

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' Routine. review of this area identified no noteworthy findings. - 0ne inspector i

follow item regarding an upgrade to Section 11.2, Liquid Waste Management 0;

. Systems,. of the updated final: safety analysis; report was closed (Section 4.2).-

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~ Maintenance / Surveillance q

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tin-the maintenance area, one unresolved item was identified regarding licensee

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submittal'to NRC of requests for. relief.from.the requirements of'the American

-Society of. Mechanical-Engineers Boiler.and Pressure = Vessel-Code, as it. applies;

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to temporary' repairs to service water systems.. The acceptability ~of the=lic-

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ensee position Jon this issue. is unresolved pending-formal' generic guidance from

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NRC (Section 5.l' 1).

An. inspection of maintenance practices concerning GE Type AKF-2-25 circuit -

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. breakers was performed pursuant to Region IcTemporary Instruction RI-86-02, l

No inadequacies were identified (Section' 5.2).

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'In.-the surveillance arca", one unresolved item concerning communication of.

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ineservice test program requirements to operations department personnel was

clored (Section 5.4.1)..

Security.

. Licensee response'to a fitness for duty program reportable event was reviewed

'withTr'eference to the' requirements of 10 CFR Part 26 and licensee procedures.

No:.inadequacie's were identified (Section 6.1).

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' Engineering / Technical Support

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y The identification by the licensee of the vulnerability of various systems-t

Limportant to: safety t'o a postulated house heating steam pipe break-is consid-

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.ered to be significant and of potential-generic interest.

Licensee response to 3[

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the finding appears to be comprehensive, appropriate, and indicative of a strength in this area.

This -item is unresolved' pending completion of licensee (tech'nical andLsafety evaluations.and the implementation of corrective action

'.(Section 7.2).

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f The control-of temporary modifications to equipment important to safety was t-reviewed within the context of NRC Information Notice 89-81. 'The inspector-considered the control of temporary modifications at Unit 1 to be adequate land

noted the apparent success of the licensee program to minimize the number of i

temporary modifications in-the plant (Section 7.3).

Safety Assessment / Quality Verification One. unresolved item was identified regarding the accident loading capability.

of the emergency gas turbine generator. An expedited load study by the lic-ensee, in response to a similar concern detailed in Region I Inspection 1 Report 50-245/90-05, resulted in the' discovery that postulated emergency loads significantly exceeded the values. described in the updated final safety analy-sis report to which the machine had been tested. A feedwater coolant injection system load was temporarily removed from gas turbine loading and the turbine was declared operable.

The issue of gas turbine generator emergency loads is unresolved pend.ing implementation of permanent corrective actions (Section 8.1,1).

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TABLE OF CONTENTS V

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1.0 Persons-Contacted...................................................

2.0 Summa ry o f Fa ci l i ty Ac t i v i t i e s.....................................

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3.0 Plant Operations (IP. 71707/71710/62703*)..........................

3.1 Control Room Observations....................................

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3.2. Plant Tours.................................................

3.3 Review of. Plant Incident Reports.......-....................

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3.4J Service Water / Condensate Systems 0utage.....................

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4.0 Radiological Controls (IP 71707/92701)...........................

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4.1 Posting-and-Control of Radiological' Areas...................

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4.2 Previously Identified Items.................................

4.2.1 (Closed) Inspector Follow Item (IFI) 50-245/89-15-01

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d Update of Final Safety-Analysis Report...............

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,5.0' Maintenance / Surveillance:(IP 62703/62705/61726/92701)...........

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5.1 Observation of ' Mai ntenance Acti viti e s.-......................... = 5

'l 5.1.1 Temporary Repair of Service Water Valve j

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'l-SW-7A~...........................................

L 5.-2 LRegion I Temporary Instruction No'. RI-86-02,

-Inspection of General * Electric Type AK-F_-2-25 Breakers......

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5'3-Observation.of. Surveillance Activities......................

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5.4-Previously Identified-. Items.................................

5.4.1-(Closed) Unresolved' Item 50-245/89-27-01, Communication of In Service: Test Program

e Requirements...................,...................

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6.0: Security.(IP 71707)................................................

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6.1 Fitness for Duty Program Significant Event..................

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Il 7.0 Engineering / Technical Support (IP 37700/93702/92703)............

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7.1 Local Power Range Monitor Spiking (Update).................

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7.2 House Heating Steamline Break Concerns.....................

'l 7.2.1 Background..........................................

7.2.2 ' Licensee Evaluation of - Safety Consequences..........

7. 2 3 Li ce n see Re s pon s e.................................... - 13 7.2.4 Conclusion.............................

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7.3. NRC Bulletin / Generic Letter /Information Notice Followup....

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f ;E ( I P. _ 92700/93702/9n! n y......................

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,<i-8il On-Site. Followup of Events...................

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Wi.M 8.1'1 : Unusual: Event - Gas Turbine Generator-l

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. Technical. Specification Limit.ing :

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Condition for: Operation Exceeded.........#,.........-....

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-8. 2 ' P e ri od i c Re p o rt s ;. -................... -........................ 1 16 :

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8.3< Plant Operations Review Committee..........'....:.............

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9.0 Management Meetings,(IP/ 30703)......................:............'..

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• The'NRC Inspection Manual Inspection Procedure (IP) or Temporary Instruction; I

s(TI) that was'used as-inspection' guidance is:1.sted for each applicable, report:

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DETAILS I

1.0' Persons' Contacted Within this report period, interviews and discussions were conducted with

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members of Northeast Nuclear Energy Company (NNECO) management and staff as necessary to support inspection activity.

'2.0 Summary of Facility Activities On April 2, 1990, at 9:53 p.m., with Mi'Istone Nuclear Power Station Unit 1 (Millstone 1 or the plant) at 100 per:cr.t of rated power, the licensee discovered a steam leak in the steam jet air. ejector room'of the turbine building.. Reactor power was teduced immeaiately to 55 percent to facili-tate attempts to isolate the leak. After determining that the leak was unisolable with the unit generator on line, the licensee commenced a nor-mal plant shutdown, with cold shutdown conditions being attained by 1:20 p.m. on April 3.

The plant remained shutdown for repairs to the service water and condensate systems until 5:41 a.m., April 7, at which time reactor startup commenced. Power ascension wu delayed for repairs to a leaking reactor head vent solenold va h e and a moisture separator level controller,~ with full power operation oeing achieved at 1:25 p.m. on -

April 9.

On. April 21, at 4:05 p.m.,

power was reduced to 70 percent as a result of high conductivity in the

"B" main condenser.

No condenser tube

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leaks were found, and at 2:19 a.m., after conductivity returned to normal, reactor power was restored to 100 percent.

Full power operation continued j

k until 9:29_a.m. on May 2, when loss of power to the hydrogen and stator cooling panel' caused a main generator runback and reactor power reduction L

to 82 percent.

Full power was restored at 11:30 a.m. and continued?until-8:30 p.m., May 12 at which time the licensee declared an unusual ~ Event

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emergency classification relating to operability of the' gas turbine gener-ator and commenced an orderly reactor shutdown as required.by technical specifications. At 9:42 p.m.,

after declaring the-gas turbine generator operable'and terminating the unusual event, power reduction was stopped at 73 percent of rated power.

By 12:50 a.m. on May 13, full power ope. ration was restored'and continued for the remainder of the inspection period.

NRC Activities The resident inspection. activities during this report period included 120

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hours of. inspection during normal working hours.

In addition, routine

review of plant operations was conducted during periods of backshifts

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(evening shifts) and deep backshif ts (weekends, holidays, and midnight

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shifts).

Inspection coverage was provided for 19.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> during backshifts-and 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> during deep backshifts.

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L A mid-SALP (systematic assessment.of license 0 performance) cycle marage-

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ment meeting was held St NRC Region I on May 8, 1990, to discuss the

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results of an inspection conducted between March 12-14.and 19-21, 1990, Results of that inspection are documented in Inspection Report 50-245/

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90-80.

a-3.0 Plant Operations E - t, 3.1 Control Room Observations c

Contrcl room instruments were observed for correhtivn between

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channels, proper functioning, and conformance.with technics 1

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specifications.- Alarm conditions in effect and alarms received in I

the control room were discussed with operators.

The inspector-periodically reviewed the night order log, tagout log, plant incident report log, key log, and bypass jumper log.

Each of the

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respective logs was discussed with operation department staff.

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inadequacies were noted.

3.2. Plant Toup I

The inspector observed plant operations during regular and backshift-tours of.the following areas:

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Controi-Room Reactor Building f

Main Battery Rooms Diesel Generator Room

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,i Gas Turbine Building Intake Structure

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Turbine Building Cable Vault i

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During plant tours, logs and records were reviewed te ensure compliance with station procedures,=to determine if entries were

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a correctly made, and to verify correct communication end equipment h

status.

No inadequacies were observed.

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3.3 Review of Plant incident Reports

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i Millstone 1 plant incident reports (PIRs) were reviewed during the

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i, inspection period to' accomplish the following:

(i) determine the I

significt. ace of the events; (ii) review licensee evaluation of the-P,

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events; (iii) verify that the licensee's response-and corrective actions were adequate; and.(iv) verify that the licensee reported the o

V" events in accordance with applicable requirements,

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The fo' lowing.PIRs warranted inspector followup and are divussed in

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the inspection report sections cited below.

• 1-90-30. Steam Leak Under SJAE Room (Sectiun 3.4)

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1-90-40,1-SW-7A Valve Bc,dy Crack (Section 5.1.1)'

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1-90-45, Potential High Energy Line Break Via House

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Heating Steam Lines (Section 7.2)

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-1-90-49, Gas Turbine LCO Determined to be Exceeded

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(Section 8.1.1)

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t 3.4 Service Water / Condensate Systems Outage

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On April 2, 1990, at 9:53 p.m., during the course of his routine tour, a' plant equipment operator discovered steam issuing-from the turbine _ building steam jet air ejector (SJAE) room into the "10" main condensate pump area.

By 11:50 p.m., licensee operators determined that the leak was not isolable with the unit generator on-line and

- h commenced a normal reactor shutdown.

Pur$uant to 10 CFR 50.72, at W

12:25 a.m., April 3, the licensee notified the,NRC of the condition via the emergency notification system. The main generator was re-moved,from the grid at 1:46 a.m. and the leak was stopped by isolat-i_ng.the main. steam supply to the air ejectors at 3:15 a.m.

Cold shutdown conditions were achieved at 1:20 p.m.

Upon entering the

SJAE: room, the source of the steam leak was determined to be a failed

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3/8-inch center plug of a Swagelock test _ fitting. The licensee de-(

cided to remain in cold shutdown to perform repairs to unit service i

water (SW), condensate, and main steam system components.

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Planned maintenance to the SW system included removal and cleaning of the rotating basket strainer, which had been experiencing relatively

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high differential pressures during normal operation.

System isola-=

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tion valve work, including replacement of a cracked yoke on the

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Valve 1-SW-26, and the SW isolation to the emergency diesel; genera-

tor, entailed draining large portions of the SW system.

In the con-i densate system, the licensee planned to re-install a new disc and

. stem into Valve 1-CN-10, the suction side isolation valve for "1C"

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main condensate pump.

This required draining of the main-condenser

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hotwell. Thus, the planned activities would remove one source of

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.on-site emergency power, and two normal methods of reactor decay heat

removal.

The' inspector reviewed the licensee's-informal operating guidelines

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established for the evolutions and observed that the service water

system lineups envisioned did not appear to be reflected in either the i

' updated final safety analysis report-(USFAR) or normal operating

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procedures, in that:

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The reactor building and turbine building closed cooling water

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systems, (RBCCW and TBCCW, respectively)_would be cross-t connected in accordance with operating procedure OP 3090, s

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" Reactor Building Closed Cooling Water System," with cooling

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-supplied by train "B" of the emergency service water (ESW)

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system. However, a caution note in Operating Procedure OP-322,

" Emergency Service Water," states that the ESW -cross-tic pro-vides cooling for RBCCW only, and that alternate cooling must be-

lined up for TBCCW and the turbine building secondary component cooling water (TBSCCW) systems, as required.

Cooling for the TBSCCW system would be supplied via a fire hose

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connection from a "B" ESW system high point vent to a-special heat exchanger end bell.

This arrangement is not governed by H

existing _ licensee procedures.

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Use of the "B" ESW system to carry plant equipment heat loads

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H during a SW system outage is discussed in the VFSAR, Section 9.2.1.2, Service Water Systems, only in the context of maintaining RBCCW system loads.

i il The inspector discussed these concerns with the operations department L

manager. The licensee responded by issuing a change to Procedure y

OP-309B, " Turbine DJilding Secondary Closed Cooling Water System,"

providing guidance for providing ESW system cooling:to the TB$CCW ~

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system.

Special procedure 90 1-02, " Service Water / Condensate Outage Guideline," was reviewed by the plant operations review committee and.

F issued on-April'4. This provided to unit operators a formal, step-L by-step sequence for the transition from normal to backup cooling h'<

supplies, including verification. that reactor temperature could be L

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controlled.

The inspector reviewed the safety evaluation which ac-I companied SP 90-1-02 and concluded the proposed system configurations

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were adequate to maintain reactor temperature within cold shutdown i

limits.

On April 7, at 5:41 a.m., reactor startup commenced, with criticality

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.being reached at 8:05 a.m.

The main turbine was placed on the grid at 2:04. p.m., on April 8, and 100 percent of full rated power achieved

[3 at 1:25 p.m., April 9.-

The inspector observed portions of the reactor

- startup,_ verified proper drywell conditions and emergency core cool-

ing system lineup, and had no questions.

b Daily unit staff meetings were held by the, licensee during the H.

outage to review progress and~ plan future activities.- Good inter-departmental communications and effective management in_volvement-during

the outages.resulted in. timely resolution of problems.

In the control f

room, the inspector observed licensee operator involvement in the il control of plant conditions, including plant restoration, startup and

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heatup, and adherence to technical specification requirements.

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Operations control during the observed evolutions was effective and-(

was considered by the inspector to be a strength, f

'4.0,._ Radiological Controls i

4.1 Posting and Control of Radiological Areas

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During plant tours, posting of contaminated, high airborne radiation, u

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E and high radiation areas was reviewed with respect to boundary identification, locking requirements, and appropriate hold points.

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The inspector identified no inadequacies.-

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m 4.2-Previously Identified Items

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. Inspector Follow Item (IFI) 50-245/89-15-01:

Update of final Safety Analysis Report J.

Thi item

umented a licensee commitment to up d

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Sei don 1

.4, " Liquid Waste Management Systems," grade

of the

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.upd6 e inal safety analysis report (UFSAR) to reflect the ; c,allation of a. vendor-operated filtration and

-demineralization skid for use in processing low purity

wastes. The inspector reviewed change 10 to the VFSAR, i

r dated March.1990, and verified that this commitment had -

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been met.

This item is closed.

l 5.0 Maintenance / Surveillance S J -Observation of Maintenance Activities The inspector observed'and reviewed selected-portions nf preventive l

and corrective maintenance to verify compliance with regulations, J

o use of administrative'and maintenance procedures, compliance with E

- codes and standards, proper QA/QC involvement, use of bypass jumpers

Q; and stfety tags, personnel protection, and equipment alignment and

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retest.

The following automated work orders were included:

' M1-89-13632; Replace 1-SW-8A after shop hydro

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M1-90-03526; Fabricate / install encapsulation for 1-SW-7A i

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M1-90-03022;. Install ESW end bell on "B" TBSCCW heat exchanger

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M1-90-02784; Replace yoke on 1-SW-26

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M1-90-02978; Implement PDCE 1-89-085 (LPRM cards)

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M1-90-Of.606; Diesel generator service water strainer drain

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modification M1-89-07401; Reassemble IC main condensate pump suction valve J

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a M1-89-03063; Replace diaphragm on 1-CU-2A

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M1-90-02998; Steam to SJAE drain level switch LS-1-16

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L 5.1.1 Temporary Repair of Service Water Valve 1-SW-7A

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On April 16, 1990, with the plant operating at 100 percent of I

full rated power, the licensee discovered a service water y?!

leak from a 90 degree circumferential' crack on the ligament

bridging the downstream flange and body of service water

t" Valve 1-SW-7A.

The leakage was' stopped by shutting the

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valve and draining the "A" turbine building secondary component cooling water (TBSCCW) heat exchanger. The'

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licensee documented the leak by initiating plant incident:

Report No. 1-90-40.

Since replacement of the valve would l

mi require shutdown of the plant to isolate service water to the valve, the licensee fabricated and installed, under

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automated Work Order M1-90-03$26, an enclosure whic6 N11y encapsulated the valve.

The temporary repair was cc.npleted on April 29 and successfully pressure tested on April 30.

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The TBSCCW system provides cooling water to auxiliary equipment in the reactor and turbine buildings during normal operations, plant shutdown, and accident conditions.

Components essential for accident mitigation and recovery

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serviced by the TBSCCW system include the reactor feed pump L

and condensate booster pump oil coolers, condensate pump-

. motors, and space coolers in the feedwater coolant injection system and emergency diesel generator areas.

The SW system is designed for 150 psig at 100 degrees F and 100 psig at 75 degrees F.

Normal operating pressure is approximately C7

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psig.

One of two heat exchangers is capable of supplying

all accident condition critical heat loads.

g Licensee procedure ACP-QA-2.18, "ASME Section XI Repair /

Replacement Program," Revision 2, states that " restoration maintenance," including encapsulation of safety Class 3.

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components, does not require a repair or replacement plar pursuant to Section XI of the ASME Boiler and Pressure Vessel Code (the Code).

However, these-repairs require.

that an engineering evaluation be performed via a non-conformance report, and that corrective provisions which constitute a system modification be processed by a plant design change request (PDCR), including review pursuant to 10 CFR Part 50.59.- The inspector reviewed the technical and seismic qualification evaluations of PDCR 1-4-90, 1-SW-7A encapsulation, dated April 26~, 1990, and concluded that all relevant issues, including affects on pipe support dead weight stress, seismic interactions,

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and heat exchanger nozzle loading' limits were addressed adequately.

Further, the enclosure entailed a pressure vessel fabricated to meet the requirements of Section VIII of the Code.

Design pressure for the enclosure was'50 psig.

The inspector had no further technical questions regarding e

this interim repair.

Section XI Article IWA-4000 of the Code specifies accept-able weld repair methods designed to restore flawed piping to original structural integrity.

Replacement of compon-ents is governed by Article IWA-7000 of Section XI.

Pur-suant' to 10 CFR Part 50.55a(g)(6)(1), the NRC may grant relief from Code requirements in order to preclude unneces-sary plant shutdown.

On a case-by-case basis, relief may i

be granted by the NRC upon licensee request.

In a letter to the Staff regarding temporary repairs to service water system piping, dated April 20, 1990, the licensee detailed its program for documenting, evaluating, and repairing defects in system pressure retaining components.

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Durlag telephone conversations between the licensee and.NRC

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on April 27-and 30, the licensee stated its position that the. encapsulation of Valve 1-SW-7A constituted a fully a

reviewed modification, that the repair met the intent of

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Section XI of the Code, and that a relief request was not

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P required.

The licensee was_ informed of the NRC position

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that the encapsulation was an. engineered temporary repair which did not reflect the system description in the Unit 1 i

Final Safety Analysis Report, and that specific relief from

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Code requirements was necessary, j

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The licensee presented its request for relief from

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L C Section XI requirements;in a letter to the Staff-on May 7.-

l Ir its submittal, the licensee stated that absent formal'

L generic guidance from the NRC on service water system re -

pairs, future repairs would be performed consistent with

L the April 20 letter without requesting further relief from the requirements.of the Code. -The inspector considered this issue to be unresolved pending issuance of further NRC

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guidance (50-245/90-07-01).

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5.2 Region I Temporary Instruction No. RI-86-02, Inspection of General

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L E1ectric Type AK-F-2-fi Tireak~ers l-The purpose of this temporary instruction was to identify the use F F of GE Type AK-F-2-25 circuit breakers in functions important to'

safety at nuclear power plants and to determine the-operating i

histories of these breakers.

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' At Millstone 1, the circuit breakers are used as field breakers for y

the emergene,v diesel generator, gas turbine generator, and reactor

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recirculation pump motor generator (MG) sets.

Regarding the diesel

generator and gas turbine generator, the breakers perform no

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safety-related functions.

For the recirculation pump MG sets, the breakers provide

afety-related trip of the recirculation pumps:

(ATWS-RPT) n a means of making the reactor suberitical diverse from

the _ reactor trip system scram.

The function is accomplished by

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energizing a breaker shunt trip coil in response to high reactor

pressure or low reactor water level signals.

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The licensee removes and overhauls the breakers during every 18-nonth'

refueling outage using Maintenance Procedure MP-772.7, "AKF 2-25 Low Voltage Power Circuit Breaker," Revision 0, dated April 27, 1989.

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The inspector reviewed this procedure and noted that it includes j

specific guidance concerning shunt trip coil replacement and moveable

and stationary contact' maintenance.

The procedure also incorporates

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the latest vendor guidance regarding proper breaker adjustment and

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lubrication, GE service information letter (SIL) 448, " Maintenance and Lubricants for GE Type AK Circuit Breaker *," dated December 23, 1986.. The licensee has ordered additional 'areakers av.d plans to

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rotate MG set field breakers every refueling outage to retuce wear

and provide a spare breaker.

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iD The ATWS-RPT trip function is tested every 18 months pursuant to

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Surveillance Procedure SP-672.1, " Manual ATWS Functional Test,"

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L Revision 1, dated August 9, 1989. The test inserts manual ATWS

Division 1 and 2 signals and verifies that the field breakers trip.

r The inspector determined that this satisfies the test requirements of NRC Generic Letter 83-28 and licensee. commitments pursuant to an l

NRC: order regarding installation ard testing of ATWS systems.

Throuch revie, vi sne licensee production maintenance management a tem records and interviews with production test and maintenance department personnel, the inspector d6termined that there were no n

1, documented failures of this type breaker at Millstone 1.

The inspectorL

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concluded that the licensee has implemented a successful maintenance

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program for these breakers and had no further questions.

This item

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is closed.

5.3' Observation of Surveillance Activities j

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Through-observation and data review of surveillance tests the

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inspector assessed licensee performance in accordance with approved

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procedures and technical specification _ limiting conditions for

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operation, removal and restoration of equipment ana review and i

I resolution of deficiencies.

The following tests were reviewed:

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SP-623.19, " Emergency Service Water System Operational Readiness Test," Revision 5-

SP-404H, "APRM Pre-Startup Functional Check," Revision 0

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L SP-633.1, " Temperature Logging During and Subsequent to Reactor j

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Heatup and Cooldown," Revision 6

SP-609.1, " Manual Scram Functional Test," Revision 4

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5.4 -Previously I*dentified Items

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5.4.1 (Closed) Unresolved Item 50-245/89-27-01 Communication

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of In-Service Test Program Requirements l

This item involved the lack of communication between the l

operations and engineering departments regarding implementa--

tion of new in-service test program requirements reflecting a

the licensee's commitment to NRC Generic Letter (GL)' 89-04.

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In response to the inspector's concerns, the licensee

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promulgated Operations Department Instruction No.1-0PS-9.03, j!

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" Guidance on the Interpretation of Surveillance Test Data for.

Pumps and Valves," dated April 9,1990.

The inspector reviewed

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the instruction and verified that the requirements of GL 89-04

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.had been incorporated adequately. Through review of the

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operations department required reading book and discussions

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p with control room personnel, the inspector was satisfied that

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licensed operators had read.and undetstood the instruction.

The inspector had no further questions.

This item is clos".d.

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' 6.0 Security

h Selected aspects of site security were verified to be proper during

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d inspection tours, including site access' controls, personnel searches, personnel monitoring, placement of physical barriers, compensatory y

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E, measures, guard force staffing, and response to alarms and degraded

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jy conditions.

No inadequacies were identified.

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6 '.1 Fitness for Duty Program _Significant Event U

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The licensee notified the resident inspector at 8:30 a.m. on

April 19,- 1990, nf a Ntnas for Duty (FFD) program reportable event.

.

A Burnes security supervisor appeared to be under the influence of j

E alcohol when reporting to work starting on the mid shift on April 19,

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1990..The individual was tested under the licensee FFD procedures and was found to have a blood alcohol content in excess of the cutoff limit of 0.04 percent. The supervisor did not assume scheduled duties, his-access was denied, and he was suspended pending further review.

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The incident was considered significant since it involved a supervisory I

individual with unescorted access to the protected area who was

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determined to be unfit for scheduled work.

The licensee reported the

_

event to the NRC Duty Officer per 10 CFR 26.73 as a significant event, i

The resident, inspector followed the licensee's reviews and corrective

actions.

In accordance with station procedures, the~ termination of

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the individual's access to the; protected area will remain in effect

for at least one year. When questioned, the individual stated that F

he had drunk alcohol earlier in the. day on April 19, but had abstained

from consumption for the'five-hour ~ period prior to reporting to work as required by NV policy.

The licensee stated that actions would be taken to clarify that the NU policy is that designated personnel must

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both abstain for five hours prior to starting duty-and be fit for duty

' when reporting for work; and further, that abstinence for the required

five-hour period is not alone sufficient to assure that one is fit for duty. _Further action relative to the guard is pending final review by

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i the contract security organization.

The inspector identified no inadequacies regarding the licensee's actions or followup to this event.

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i 7.0 Engineering /hechnicalSupport

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7.1 Local Power Range Monitor Spiking (Update)

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h Spiking' of local power range monitors (LPRMs) associated with _

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certain General Electric Company (GE) Model NA-300 fission chamber i

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P detectors was detailed in Region ! Inspection Report 50-245/89-25,

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L Section 5.2.2.

In some cases reactor protection system half-scram-

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actuations have occurred on average power range monitor (APRM)

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channels associated with these LPRMs.

Thus, the phenomenon could

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increase the potential for direct reactor trips, and trips during

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maintenance and. surveillance activities.

As a long-term solution to the problem GE has developed prototype

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LPRM first stage amplifier cards designed to eliminate the current

spikes while maintaining the design time response characteristics of

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the system.

The new cards are classified as QA Category 1, but do j

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not meet all existing seismic criteria.

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On April 25, 1990, under automated Work Order M1-90-02978,' the

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licensee. installed eight prototype cards into the LPRM groups not associated with APRMs.- After successful testing of these cards,

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c fully qualified' production cards will be installed into the APRM i

i channels.

The. inspector reviewed Licensee Safety Evaluation

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No. ISE/MP3-89-097, Revision 0, dated November 20, 1989, associated

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with plant, design, change Evaluation MP-1-89-085, under which the cards were installed.

The rod block monitor.(RBM)'is the only

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safety equipment affected by the modification.

The RBM system is designed to prevent ex m ding fuel Germal' limits due to individual L

control rod withdrawal at power by inserting rod blocks if LPRM average power' exceeds a flow-biased APRM reference value.. The c

licensee determined that since only Groups 1 and 2 LPRMs are.

affected by the modification, the RBM remains capable of mitigati_ng

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a rod withdrawal error event.

Based on deterministic analysis by GE,-

E Lthe licensee also concluded that the cards do not degrade the existing e

seismic margins for the affected control panel, and that the cards can withstand any credible seismic environments in the-panel.

Finally, the licensee noted that the RBM is not credited or required for mitigation of or recovery from a seismic event..The inspector agreed that the modification presented no unviewed safety question pursuant to 10 CFR Part 59.

The ' inspector reviewed the LPRM spiking log maintained in the control room and noted_that the incidence of spikes from the Group 1 and 2 LPRMs appeared to have decreased substantially since the installation

of the prototype cards.

Final resolution of this issue will be followed during future routine inspections.

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7.2 House Heatino Steamline Break Concerns

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7.2.1'

Bac kor'ound.

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On May 2, 1990, the licensee informed the inspector that.

due to concerns regarding the effects on equipment' import-ant to, safety of & postulated break of house heating (HH)-

steam lines in the turbine building HH steam to-Millstone

L I had been isolated.

TheLsteam lines affected. included a

h 4-inch header and a 10-inch header in the _switchgear and h

heating and ventilation (HVAC) areas, respectively, of the turbine building. On May 9, the inspector was. briefed on-i this.1ssue by the Millstone 1 director and the acting en-T gineering department manager.

The licensee identified L

these concerns as a result of evaluations'and system walk-

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downs performed to support a plant design change to replace heating and cooling coils on switchgear area supply. fans

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HVS-6A and 6B.

Pursuant to its administrative procedures, if the licensee initiated a reportability/ operability evalua--

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tion of this condition on May 1.

Criterion 4 of 10 CFR Part 50,- Appendix A requires, in part, o

that systems, structures, and component $'important to safety shall.be designed to accommodate'the effects of, and to be

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' compatible with, the environmental condition associated with normal operation and postulated accidents. This equipment h

shall be protected from the effects of discharging fluids which may result from equipment failures.

Millstone 1 Updated Final Safety Analysis Report (VFSAR),

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Seption 3.11.1, identifies the equipment required to func-

tion during and after postulated accident conditions, in-

cluding rupture of high energy lines.

Equipment located

in the areas of concern include the standby gas treatment, i-control room ventilation, AC emergency power, and DC emer-gency power systems.

Non-IE class equipment includes the-t reactor protection system motor generator sets.

Section l

3.11.1.1.6 of the VFSAR describes the environmentally. con-

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trolled areas of the plant provided to protect equipment important to safety from the harsh environments postulated,

in part, by a high energy line break (HELB).

These " mild

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environment" areas include the turbine building switchgear

'7 room (mezzanine floor at elevation 34'6"), and the HVAC

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equipment room.

7.2.2 Licensee Evaluation of Safety Consequences I

i The inspector reviewed the licensee's preliminary technical'

evaluation to determine the scope of the safety concerns involved.

Of the several issues identified by the licensee,

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the 4-inch HH steam line in the area housing the essential

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H 4160VAC and 480 VAC switchgear. With steam being supplied-L'

to Millstone 1 from the Millstone 2 reboiler, no automatic.

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trip of the steam supply would be likely to occur.: Conser-vatively assuming no escape for the steam and taking.no..

b credit for condensation, th; i;c see concluded that there p

was reasonable doubt of the ability of a majority of the-electrical equipment concerned to survive the transient, f

TMs equipment includes:

station battery chargers

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vital-and instrument motor generator sets

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reactor protection system motor generator se+,

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.both divisions of safety related AC distribution

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Fc equipment, including power supplies to emergency (,

core cooling systems L

'The resulting condition equates to a station ' blackout (SBO)'

with-little chance of restoring a normal AC power configura-

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tion within the assumed SB0 restoration period of four hours.

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During a walkdown of the turbine building HVAC room on April 26, the licensee also identified the following con-

cerns:

j Redundant switchgear area supply fans HVS-6A and 6B-

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may be susceptible to a common mode failure of the upstream. steam heating coils.

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The, fans are not environmentally qualified.(EQ).

No drains exist.in the common discharge' plenum of

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HVS-6 to preclude carryover of steam / water into the

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switchgear. room.

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The HVAC room is classified as an EQ mild environment.

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However, a 10-inch HH steam line and smaller branch.

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lines are located above and adjacent to non-EQ motor control centers (which supply HVS-6), the unit's main

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fan control panel, control room ventilation system; i

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components and the. standby gas treatment system.

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The seismic qualification of the switchgear area HVAC

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system may not be documented adequately.

A: Northeast Utilities calculation of the heat gain in the'

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switchgear area du.%:a the summer period indicated that,

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with the gas' turbix generator operating during a' loss of.

i offsite power event, ambient temperature would increase to 104' degrees F (equioment design temperature) in approxi-

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mately 30 minutes.

As reported in NRC integrated plant safety assessment report, NUREG-0824, Supplement 1, dated-

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i November 1985, Millstone 1 is committed to operate the switchgear area supply and exhaust fans if this event occurs. ' Failure of the fans to operate could affect ad-

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versely the ability of safety. systems to function.

7. 2. 3'

Liceesee Response Ba',ed on its preliminary findings, the licensee placed an

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administrative hold on use of the HH system in the affected

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areas of the unit. On May ll, pursuant to 10 CFR 50.72(b)

(2)(iii)(A) and (B), the licensee notified the NRC of-the

ondition. An investigation was initiated to determine the i

potential for similar vulnerabilities at Millstone 2 and 3

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and the Haddam Neck nuclear power stations. A comprehensive

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review of the safety consequences of a high energy 'line

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break of the HH system, and the seismic qualification of

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certain HVAC systems is ongoing.

7.2.4 Conclusion

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The inspector found the licensee's response to the issues reviewed to be prompt, comprehensive, and indicative of a high regard for plant safety.

The licensee's timely notification of this potentially generic issue to the NRC

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was considered by the inspector to be appropriate.

The i

engineering and technical resources applied by the' licensee to identify and evaluate the consequences of a HH' steam line rupture are noteworthy.

The inspector concluded that the licensee's actions are consistent with continued safe i

operation of the plant and had no further questions.

However, dup to the ongoing licensee evaluations and potentially generic nature of the HH steam issue, this is considered to be an unresolved item (50-245/90-07-02).

7.3 NRC Bulletin / Generic Letter /Information Notice Followup On December 6, 1989, the NRC issued Information Notice (IN) 89-81,

. Inadequate Control of Temporary Modifications to Safety-Related

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' Systems.

The notice alerted licensees to potential problems re-sulting from inadequate control of temporary modifications to safety-related equipment.

Specific items of concern included im-proper tracking of temporary modifications, inadequate or untimely technical and/or safety reviews, and failure to ensure that control

room drawings reflected the modifications.

At Millstone 1, the governing administrative procedure for temporary modifications is ACP-QA-2.068, " Station Bypass Jumper Control,"

Revision 9, dated October 6, 1989.

Station policy is to limit the number and duration of jumper devices installed in the plant.

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general, modifications known to require installation beyond three months are processed as plant design change '

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Installatiod The procedure requins that temporary modifications l

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to operable equipment receive techni al and safety assessments to

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identify possible adverse effects on. safety, reliability, or

efficiency. Mere identified, techni:al and safety evaluations are

.then performed and approved by the plant operations review committee l

(PORC) prior to implementation of the modification.

For inoperable

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equipment, the PORC must review and alprove the. modification within i

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.34 days of installation.

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Review and Audit: A record of installed temporary modifications p

n (jumper devices) is maintained by the shift supervisor in the

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control = room. The log is reviewed by the shift supervisor each

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L shift and-prior to changes in plant operating mode.

The operations k

department manager audits the log monthly to assess the need for the

modifications and to identify those installed for greater than three i

months.

The responsible department head and the PORC review all-

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modifications installed for greater than three months and provide a t

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written determination that the modification is still needed or.

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whether the change should be made permanent.

Temporary modifications.

i installed for greater than six months must be reviewed by the unit

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director with the Millstone station director.. Finally,-the vice president;of nuclear operations is informed by memorandum of'

t the status of these modifications.

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Drawings: Jumper devices installed for greater than seven days are

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reflected in applicable control room drawings by attaching-a copy of

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the device control sheet to the drawing.

The' inspector reviewed selected jumper devices and log sheets and I

l verified that licensee administrative requirements were being a

satisfied. The inspector noted that only one safety-related system was affected at the time of the inspection.

The inspector concluded that adequate licensee controls exist to document, track, review,:and'

E audit temporary modifications of safety-related equipment at -

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. Millstone 1.

This item is closed.

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8.0- Safety Assessment / Quality Verification

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8.1 On-Site Followup of Events

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8.1.1 Unusual Event - Gas Turbine Generator Technical

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Specification Limiting Condition for Operation Exceeded

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On May 12, 1990, at 8:30 p.m., with Millstone 1 at 100 percent of full rated power, the licensee determined that

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the gas turbine generator (G/T) did not have sufficient capacity to support postulated accident loads.

The G/T was i

declared inoperable and an orderly plant shutdown was com-

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menced pursuant to technical specification requirements.

In accordance with licensee emergency plan implementing

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procedures, an Unusual Event emergency classification was declared.

The NRC was notified of the event via the emer-gency notification system as required by 10 CFR

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50.72(b)(1)(1)(A), initiation of any nuclear plant shutdown required by plant technical specifications.. After review by the plant operations review committee (PORC), the acci-dent loads on the G/T were reduced by inhibiting the automatic initiation of the feedwater coolant injection system (FWCI), thus p' cing the accident loading on the machine within its proven capacity.,

The FWCI system was declared inoperable, the G/T declared operable, and the reactor power reduction secured by 9: 42 p.m.

The Unusual

. Event was terminated at 9:54 p.m. and full power operation restored by 12:50 a.m., May 13.

The G/T is one of two emergency power sources designed to start automatically on a loss of normal offsite power or initiation of the emergency core cooling system.

It sup-plies power to one division of low pressure emergency core cooling systems and the high pressure feedwater coolant injection (FWCI) system.

Technical specification 3.5.F.3,

" Minimum Core and Containment Cooling System Availability,"

limits operation with the G/T inoperable to four days, after which an orderly shutdown is required.

Table 8.3.7 of the Millstone 1 Updated Final Safety Analysis Report (UFSAR), states that the emergency load for the G/T is 9.913 MW.

Surveillance procedure 668.2, " Gas Turbine Emergency Fast Start Test," Revision 12. Cha ige 1, demon-strates the operability of the machine by loading it to greater than 10 MW, and recently the licensee had tested the G/J at approximately 10.2 MW.

As a result of previous G/T loadina concerns detailed in Section 7.1.2 of Region I Inspection Report !0-245/90-05, the licensee expedited an on going emergency load study.

The licensee's preliminary calculations revea.ed that with the FWCI system at runout, pumping relatively cold water from the condensate storage tank, the G/T emergency-load would be approximately 11.463 MW.

Table 8.3.5 of the Millstone 1 UFSAR states that the load capability of the G/T at 100 degrees F ambient temperature is 11,1 MW. Thus, the required emergency loads were greater than the proven capacity of the machine.

By removing the post-LOCA/ loss of normal power automatic initiation circuit control power fuses, the licensee reduced emergency loads below the pro-ven capacity of the machine, and entered the seven-day-action statement of Technical Specification 3.5.C.3 for the FWCI system. The inspector verified through review of cir-cuit diagrams and discussions with instrumentation and con-trols department engineers that removal of the fuses would

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not inhibit the automatic response of other safety systems

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h and that operators retained the ability to operate FWCl

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system components manually. Through telephone conversa-

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tions, the NRC staff and Region I management were kept

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infor,7.ed nf these :ctivities.

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On May 14, the inspector attended a meeting of the PORC.

The committee reviewed jumper / bypass Control No. 1-90-09,

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nE which tracked removal of the FWCl fuses and its associated l

h safety impact evaluations.

Based on the fact that the FWCI i

b system is not credited in the loss of coolant accident.

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analysis described in the UFSAR, the licensee determined that no unreviewed safety question existed.

3 The inspector concluded that the committee demonstrated a

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conservative attitude towtrd safe plant operation and that its deliberations were probing and technically sound.

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inspector had no further questions.

Because the solution r

h to the problems uncovered during this event is. ongoing, the inspector considered the issue of G/T operability to be an unresolved item (50-245/90-07-03).

8.2 Periodic Reports Upon receipt, periodic reports submitted pursuant.to technical specifications were reviewed. The inspector also ascertained s

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whether any reported information should be classified as an

abnormal occurrence. The following reports were reviewed:

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Monthly Operating Report - March 1990

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Monthly,0perating Report - April 1990

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This review verified that'the reported information was valid and-included the required NRC data.

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8.3 Safety Assessmen' ' Quality Verification The inspector attended three plant operations review committee (PORC)

' meetings during the inspection period and verified that Technical

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Specification 6.5.1 requirements for committee quorum were met.

Meeting agenda included review and approval of plant design modifi-

.

cations, setpoint change requests, procedure revisions, plant inci-dent reports, and technical specification change requests. The com-mittee discharged its functions in accordance with relevant require-ments and demonstrated through detailed and frank discussion an ap-propriate regard for nuclear safety.

No inadequacies were identi-fied.

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9.0.-Management Meetinos L

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Piriodic'meetingswereheldwithstatio'nmanagementtodis' cuss.

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inspection findings.during the inspection period.; A summary of

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- findings was:also. discussed at the, conclusion of the-inspection,'

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No proprietary information was covered within the scope of the w

gJ jinspection. - No written material was givento'theilicensee during-

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the-inspection period.

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