IR 05000321/1986030

Insp Repts 50-321/86-30 & 50-366/86-30 on 861005-09. Violations Noted:Failure to Provide or Implement Procedures & Failure to Perform Adequate Valve Stroke Time Testing

ML20214V074
Person / Time
Site:	Hatch
Issue date:	11/20/1986
From:	Stadler S, Wilson B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20214V005	List: IR 05000321/1986030 ML20214V001 ML20214V026
References
50-321-86-30, 50-366-86-30, NUDOCS 8612090515
Download: ML20214V074 (15)

Text

{{#Wiki_filter:-

.
*

M Effo UNITED ST ATES

/ o  NUCLEAR REGULATORY COMMISSION REGION 11
[  ,

101 MARIETTA STREET, g j

* I e  ATLANTA, GEORGI A 30323
'+4.....

Report Nos.: 50-321/86-30 and 50-366/86-30 Licensee: Georgia Power Company P. O. Box 4545 Atlanta, GA 30302 Docket Nos.: 50-321 and 50-366 License Nos.: DPR-57 and NPF-5 Facility Name: Hatch 1 and 2 Inspection Conducted: October 5-9, 1986 - Inspector: S.' D~.

  -

Stadler

     [[Date Signed Accompanying Personnel: H. Christensen K. Poertner G. Schnebli C. Smith Approved by: C-     /// 2.ON(

B. WilsW, Section Chief /DatetSigned Operations Branch Division of Reactor Safety

 "

SUMMARY Scope: This routine, unannounced inspection was in the area of maintenance activitie Results: Two violations were identified - Failure to provide or implement procedures and failure to perform adequate valve stroke time testin . I PDR

*

. REPORT DETAILS Persons Contacted ,

   '

Licensee Employees

*H. Nix, Plant Manager
* Seitz, Manager, Maintenance
* Fraser, Acting QA Manager
* Sumner, Manager, Operations
* Powers, Manager, Engineering
*M. Googe, Manager, Planning and Outage
*R. Baker, Manager, Nuclear Licensing
*C. Moore, Manager, Training and Emergency Preparedness
*J. Simmons, Superintendent Plant Administration
*G. Sorrell, Document Control
*S. Bethay, Regulatory Compliance
*R. Moxley, QA Audit Group
*T. Elton, Plant Engineering T. Anderson, Supervisor, DCR Implementation /SCS, In C. Patel, DCR Implementation Engineer R. Dewberry, Manager Design Change and Modification /SCS, In B. Garner, Manager, Nuclear Plant Support /SCS, In R. Glasby, Project Manager /Bechtel J. Harris, DCR Implementation Engineer K. Hughes, Design Supervisor /SCS, In B. Matthews, Supervisor of Design and Drafting /SCS, In R. Mehta, Senior Electrical Engineer /Bechtel J. Newton, Maintenance Supervisor, Work Planning Group T. Quayle, DCR Implementation Engineer A. Vora, DCR Implementation Engineer J. Wade, Plant Engineer, Reactor Systems R. Watlet, DCR Closeout Engineer Other licensee employees contacted included engineers, technicians, operators, mechanics, security force members, and office personne NRC Resident Inspector
*P. Holmes-Ray
* Attended exit interview Exit Interview The inspection scope and findings were sumnarized on October 9,1986, with those persons indicated in paragraph 1 above. The inspector described the areas inspected and discussed in detail the inspection findings listed belo No dissenting comments were received from the license The

'

.

 ^

licensee did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspectio . Licensee Action on Previous Enforcement Matters This subject was not addressed in the inspectio . Unresolved Items

- New Unresolved items were not identified during the inspectio . Maintenance Observation (62700, 62703)
- The inspectors observed ongoing maintenance activities and reviewed records
'to verify the work was conducted in accordance with approved procedures, i Technical Specifications, and applicable industry codes and standards. The inspectors also verified that: administrative controls were followed; tagouts/ clearances were adequate; radiological controls were proper; fire protection was adequate; quality control hold points were adequate and were observed; the procedures specified in the Maintenance Work Packages were adequate for the scope of the maintenance performed; required administrative approvals were obtained prior to initiating the work; the provisions for housekeeping and cleanliness were adequate; maintenance records were assembled and stored as part of the maintenance history; vendor technical manuals were used by maintenance personnel when applicable; in areas where-the maintenance activity was performed by contractor personnel the contractor was controlled in accordance with the licensee's approved quality assurance program; and licensee supervisory oversight was adequate for both GPC and contract maintenance personne Throughout the inspection period, the inspectors witnessed maintenance activities in the plant service water (PSW) building, reactor building, the diagonals, the drywell, and the torus. The activities observed included the following: . replacement of the air release valve for 2B PSW pump; trouble-shooting the control circuit for PSW pump 20 minimum flow valve; inspection of Limitorque motor operated valves for compliance with environmental qualifications; spent fuel pool sipping and inspection of fuel assemblies; preparations for local leak rate testing of containment isolation valves in the torus; 18-month and 36-month preventive maintenance activities on motor operated valves; alignments on main steam pressure transmitters in the torus; blue checks of valve 2E11-F0288 wedge and sea Prior to entry into the drywell, the inspectors were required to attend an ALARA briefing. The briefing is mandatory for all personnel prior to performing maintenance in radiation areas. The briefing was conducted by health physics technicians and used projected overlays to show the maintenance are All problem areas concerning radiation and contamination levels were discussed in detail. Low radiation level areas were identified to aid in minimizing dose. The inspectors considered the briefing to be very beneficial with potential significant impact on the ALARA program during the current Unit 2 outag .
.
.

The inspectors observed I&C technicians working on the 1B diesel generator air compressor start switch 1R43-N002C under Maintenance Work Order (MW0) 1860872 This M0W indicated that the low end switch calibration was drifting out of tolerance. Procedure 34SV-R43-001-25, Diesel Generator Air Start Test, requires that the compressor starts at 235 psig ( 5 psig), but the compressor was not starting until 227 psig. The minimum Technical Specification starting pressure is 225 psig. MWO 18608724 also stated that

"this is the third MWO since 7/30/86, work this one as MWO 1-86-7117 was not." The I&C technician indicated that he attempted to find the other two MW0s to determine what work had previously been accomplished on this switc He indicated that he could not locate the MW0s due to the licensee's methodThe of splitting-up completed work packages for record storage purpose inspectors subsequently retrieved computer printouts of the three MW0s from the automated MWO tracking syste The original MWO, number 18607117 was written on July 30, 1986, but was not signed on for work until September 16, 1986. A month and a half to start work on a safety-related switch which was known to be outside of the procedural tolerance, known to be drifting low, and known to be very near the Technical Specification limit of 225 psig, seems excessiv On August 30, 1986, another MWO number 18607919 was written on this diesel generator start switch. This MWO indicated that the compressor did not meet the required starting pressure during the air start tes The first MW0, 18607117, had specifically requested the switch to be repaired or calibrated to allow this air start test to be conducted. Had this MW0 been completed on a timely basis, the test on August 30, would have been successful and the second PWO would not have been necessary. This compressor switch was finally calibrated but not repaired under MWO 18607919 on September 6,1986. The MW0 was marked as unsatisfactory On by Operations, September 16, however, and a third MW0, number 18608724, was writte , MW0 18607117, written on July 30, 1986, was marked as no work performed due to the calibration performed on September 6. This VWO was again marked as unsatisfactory by Operations and the corrective action was the same as under MWO 18607919, to write a new MWO, number 186872 The apparent frustration expressed by Operations personnel in generating three MW0s on the same piece of equipment, and the apparent lack of timely and adequate corrective action by the Maintenance Department, were cause for concern to the inspector The inspectors observed I&C technicians working on the 2D plant service water discharge pressure gaug An MWO had been generated on the gauge because it was reading 40 psig when the pump was out of service. The techniciam, removed the gauge and attempted to calibrate it utilizing a portable test rig. They indicated to the inspectors that the gauge could not be repaired and that a new gauge would be ordered. The defective gauge was reinstalled and placed in service and the MWO cleared. The inspectors noted that no local indication that the gauge was defective, such as a problem tag or out-of-calibration sticker, was provided. A review of the I

next day of the operator log sheets provided no indication next to the pressure reading recorded that this gauge was reading high or was defective.

, l The licensee apparently does not have a formal program to ensure that defective equipment, controls, and instruments are labeled until repaire The Electrical Department utilizes informal, hand-written problem tags for l l

. _ .

'
.

this purpose, but the practice has not been extended to Maintenance and Instrumentation and Control (I&C). Tags or labels placed on defective equipment can help prevent incorrect entries on safety-related logs and operating errors, and can also prevent duplicate MW0 The inspectors reviewed maintenance procedures . and several completed maintenance work orders (MW0s) to determine if the licensee was performing maintenance activities in accordance with their governing administrative procedure Three MW0s completed in August 1986, 2-86-2136, 2-86-2137,,and 2-86-2154, pertain to the performance of the 18-month and 36-month preventive maintenance (PM) on safety-related Limitorque valve operator MW0s 2-86-2136 and 2-86-2137 were for the iesidual heat removal (RHR) service water Limitorque valve operators 2E11-F068A and 2, respectively, cooling water pressure control valves for the RHR heat exchangers and MWO 2-86-2154 was for the core spray valve 2E21-F0318. The MW0s required the PMs to be performed in accordance with PM procedure 52PM-MNT-005-0S, Limitorque Valve Operator Inspection. When these MW0s were perfonned, Revision 3 was in effect. The 36-month PM required lubricating the geared limit switch, which requires the removal of the geared limit switch from the

' limit switch component. Upon completion of the lubrication and reassembly, the limit switch was to be adjusted in accordance with maintenance procedure 52GM-MNT-017-05, Limitorque Valve Operator Setup and Test. Procedure 52GM-MNT-017-0S was replaced on February 6,1986, by general maintenance procedure 52GM-MEL-022-0S, Limitorque Valve Operator Electrical Maintenanc A review of the PM data sheets and the MWO actual work performed, indicate that the limit switch compartment lubricant was replaced, but that the general maintenance procedure for verifying the correct limit switch setup was not performed. The licensee stated that the maintenance foreman apparently made the determination that the limit switch adjustments were not needed, so procedure 52GM-MEL-022-0S was not performed. This evaluation was not documented on the PM data sheet or the MWO. At the completion of the PM, the Limitorque valve operators were cycled, and cycle time with proper valve position indication was recorde Procedure 52PM-MNT-005-05, revision 3, step 7.6.6.13 states, perform limit switch adjustment in accordance with 52GM-MNT-017-05, Limitorque Valve Operator Setup and Tes The licensee was informed that failure to follow procedures, t$ree examples, is a violation (50-366/86-30-01).

The inspectors held discussions with responsible personnel from the newly formed Outage Management Group (OMG). The group is staffed with members from GPC and contract personnel from Bechtel. The discussions centered around the processing of an MWO from initiation to completion; the inter-faces between work planning personnel, maintenance and contract personnel; and problem areas identified during the outage. The inspectors reviewed several MWO packages ready to be issued to maintenance personnel (MWO 2-86-184, 2-86-2423, and 2-85-3145) and several completed MW0s (2-86-2210, 2-86-2259, 2-86-2368, 2-86-2929, and 2-86-2613). OMG and GPC management indicated the major problem associated with the outage thus far had been the computer program for scheduling the maintenance, in that the program did not prioritize maintenance or recognize plant conditions required for the maintenance to be performe This issue was resolved by hand sorting the

'

.

MWO packages to coordinate the work. During these discussions, OMG indicated the outage was extended .to 51 days versus the 47 days initially projected due to problems with certain critical systems and the addition of approximately 500 MW0s to the schedule. The key work items associated with the critical systems included the following: reactor assembly - delayed due to refueling bridge axle repair; main steam - delayed due to a main steam isolation valve (F022B) failing local leak rate testing (LLRT); core spray - delayed due to Maglock switch replacement and pump and valve repair work; control rod drive - delayed due to refueling crane -problems; residual heat removal - delayed due to E11-F028B failing the LLRT due to a worn edge which will require replacement or major repai The inspectors reviewed the status of Unresolved Item 321,366/86-22-0 This item dealt with electrical backseating of motor operated containment isolation valves on a routine basis to prevent excessive packing leakage, and the licensee practice of stroke time testing power operated valves from light to ligh In the area of electrical backseating of containment isolation valves, the licensee has issued a special procedure, 52SP-100286-JD-1-05, Electrical Backseating with " Instantaneous" Circuit Breaker Trip Protection. This procedure installs a circuit breaker that trips at approximately twice the normal running current. The licensee expressed an-opinion that deenergizing the valve motor at approximately twice the running current is more conservative than allowing the valve to torque out on the torque switch setting. The licensee has also issued a standing order to remove all electrically backseated valves from their backseated position during a plant cooldown evolution to prevent thermal binding. The licensee took these actions as a result of a RCIC steam isolation valve that was electrically backseated and failed to close on an isolation signal during a recent plant cooldown. The inspectors reviewed the special procedure and the standing order and consider that the approach presently taken by the licensee is more conservative than the method of electrical backseating used in the past; however, the inspectors do not consider that routine electrical backseating of containment isolation valves is- an acceptable practic During this inspection, the inspectors determined that the licensee has already electrically backseated two Unit 1 containment isolation valve Both these valves had their packing replaced during the recent Unit 1 outag During the Unit 2 outage, the licensee presently plans to disassemble five containment isolation valves that were routinely electrically backseated and inspect for valve damag Until further licensee and NRC evaluation, this item will continue to be carried as unresolve During the inspection documented in Inspection Report' 50-321,366/86-22, the licensee indicated that relief had been requested from NRR to allow stroke time testing of power operated valves from light to light as opposed to initiation of the actuating signal to the end of the actuating cycle as required by ASME Section XI; how.ver, the licensee could provide no documentation to support this statemen Although Hatch site personnel requested, in letter NED-85-034,1363N, that the Inservice Inspection Program be revised to clarify testing methods used

.

   .
.
.

for stroke timing of valves,- the licensee could still not supply documentation that relief had been granted from the requirements of ASME Section XI. Review of the licensee's ISI program identified that no relief-request was submitted. Unit 1 Technical Specification 3.7.D.1 states:

 ' "During reactor- power operation, all primary containment isolation valves listed in Table 3.7-1 and all reactor coolant system instrument line excess flow check valves shall be operable except as stated in Specification 3.7.D.2."

' Unit 1 Technical Specification 4.7.D.1 states:

 "At least once per operating cycle the operable isolation valves that
      ~

are power operated and automatically initiated shall be' tested 'for simulated automatic initiation and the closure times specified in Table 3.7-1."

Unit 2 Technical Specification 3.6.3 states:

 "The primary containment isolation valves and the reactor instrumenta-tion line excess flow check valves specified in Table 3.6.3-1 shall be OPERABLE with isolation times as shown in Table 3.6.3-1."

Unit 2 Technical Specification 4.6.3.3 states:

 "The isolation time of each power operated or automatic valve specified in Table 3.6.3-1 shall be determined to be within its limit when tested pursuant to Specification 4.0.5."

Unit 2 Technical Specification 4.0.5 states:

 "Following start of facility commercial operation, inservice inspection of ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g),

except where specific written relief has been g anted by the Commission

pursuant to 10 CFR 50, Section 50.55a(g)(6)(i)

I ASME Section XI defines stroke time as the time interval from initiation of the actuating signal to the end of the actuating cycle.

l The E. I. Hatch Pump and Valve Test Plan states: The limiting value of ! full-stroke time of each power-operated valve has been specified. Full-stroke time is that time interval from initiation of the actuating signal to the end of the actuating cycle.

l .

-+r+ eMg-* ewe- - w.m-wg +--9mp--rqm r-wweg+m +gyem--- -

W+- -e +emm -+*

- - . .. . - . . - - ._  .- -. . - - - .- -

~

.

.

.

. l 7 . Contrary to the above, the licensee does not stroke time test power operated valves from initiation of the actuating signal to the end of the actuating-cycle as-required by Technical Specifications and ASME Section XI, and has not received specific written relief from the requirements of the ASME code that requires stroke time testing from initiation of actuating signal to the end of the actuating cycle. This failure to meet the requirements of the Technical Specifications and ASr4E Section XI is identified as a violation ,

 (50-321,366/86-30-02). Diesel Generator 2C Low RPM Test Emergency Diesel Generator 2C was scheduled for an extended low RPM test run due to major maintenance that was performed during the current outage. The vendor representative wrote a special procedure 52 SP-100386-1C-1-25, Diesel Generator 2C Low Speed Run, to be utilized during the conduct of the tes Special procedures at Hatch require only the department superintendent's
,  approval, in this case the Maintenance Superintendent, and no review by the i  Plant Review Group. This was considered a maintenance test so it was to be

directed by a maintenance foreman and the vendor representative with 5 assistance by Operations personnel. The System Engineer assigned to the

"

diesel generators did not appear to be directly involved in the coordination of the tes On Monday, October 6,1986, preparations had been completed for the test i when the inspectors arrived at the 2C diesel generator. These preparations j included lifting leads to defeat the low coolant pressure trip at reduced

] RPM The operators expressed a concern to the inspectors regarding the

interface between the special procedure and the normal operating procedure l utilized for local diesel generator operations. Since the special procedure

- did not contain all' the prerequisites and steps in the normal procedure, i they planned to utilize both procedures for the test. In a number of areas,

' however, the two procedures contained different requirements for starting i and loading the diesels, and the operators were not sure how they were going i to resolve these differences during the tes The resolution of the i procedure differences should have been accomplished by plant management or

the Work Planning Group prior to assigning operators to perform the tes Subsequently, the licensee made the decision to delay the diesel test until

'
;  the special procedure could be rewritten. In addition, several problems i  were encountered with valves in division 2 of plant service water (PSW)

j which provides cooling water to the 2C diesel generator. The air release valve in the discharge line of the 2D PSW pump would not reseat and was allowing excessive flow loss of PSW. Licensee investigation of the problem

. determined that a guide piece in the ball check had sheared off allowing the ball to become cocked. The valve was repaired, and a new design to be  ,

installed at a later date has the guide incorporated as an integral part of ' the valve body which should prevent additional failures of this type. The

:  other valve which failed to seat allowing a loss of PSW flow was the 28 PSW 5-  minimum flow valve. Repair of this problem required instrument and control-(I&C) technicians to reset the Fisher controller which was out of adjust-
>  men The inspectors observed the maintenance activities and documentation j  involved with these PSW valves and they appeared adequate. The special
!

_ __ __ _ . _ _ . _ _ _ _ . . _

     . . - .  - - - - - .

_

 ,
..

.

procedure rewrite and the'PSW valve repairs delayed the 2C diesel low RPM , test until Wednesday, October 8. Although the special procedure had been revised to contain requirements fran the normal procedure, all of the prerequisites had not been incorporated. This required the operators to , again utilize both procedures for the test. The inspectors observed the initial portion of this test and noted a number of deficiencies which were related to the license The vendor requested that the operators depress the voltage trip test prior to beginning'~the diesel tes This was done to prevent overheating the field at low RPMs and the trip was to be reset at full (900) RPM The trip and reset should have been included in the special procedure but were no The vendor directed the operator to momentarily depress the engine ^ start button but the diesel did not start. These particular diesels did not have the 12 second seal in start circuit the vendor apparently expecte 'The special procedure contained a " caution" directing the operators to closely monitor the diesel cooling water pressure at low RPMs because the low cooling water pressure trip had been defeated. At approxi-mately 400 RPMs, the inspector asked the operators if the cooling water pressure was adequate in accordance with this caution. There appeared to be some confusion over which gauge to read since none were specifically marked as cooling water pressure. Once it was decided that the jacket water pressure gauge was correct, the operators indicated that they did not know a minimum acceptable pressure at reduced (400) RPMs. The acceptance criteria was r.ot stipulated in the . special procedure or the caution statement. In addition, the operators did not know, and could not find in the normal procedure, a minimum i acceptable value for full RPM operation. This lack of specific criteria in the special procedure rendered this caution useless, and did not ensure adequate cooling water pressure to the diese The normal procedure utilized for local diesel operation contained a caution against operating the diesel generators at low loads for .

        ,

extended periods of time. This caution is designed to prevent fires due to an accumulation of exhaust gases at low loads. Low loads and extended period of time were not defined in the procedure, and it was not clear if this caution was applicable to these test condition At approximately 800 RPMs the vendor decided to check the crankcase oil level, perhaps due to an unusual noise level. The crankcase dipstick , indicated the oil level to be very low, at less than the add mark for ' operating conditions. The test was secured to add oil to the diese Two drums, or approximately 100 gallons, were added. The licensee indicated that the level had been incorrectly checked by the operators prior to starting the circulating pump and filling the lines and filte Although the vendor believed that the diesel was not damaged by running a 100 gallons low, the checking of the post maintenance oil .

  .

s . v-se--w e-w w --n--vm w wew-r-w w mv e mm m r-w e ,*. ear- w e e m , , - . . eew

  -  _- __
  / . ~4 1 .
*  '
.,    .$

1 level following maintenance should be. procedurally controlled to ensure it is done correctly .and to ensure emergency diesel generator availability.-

 - The diesel generator was - restarted approximately two hours later and brought to 800 RPMs innediately. At this time, the room filled with smoke and it was indicated to the inspectors that-the ventilation system had not been properly aligned prict to the tes Special procedures _ involving safety related system or equipment should be
' subject to the same quality assurance controls as normal safety related
    -

procedures including review and approval by the Plant Review Board (PRB).

Section 6.8.1.c of Hatch Technical Specifications requires that written . procedures shall be established and implemented for surveillance and test activities of safety related equipment. This failure to provide adequate procedures on two occasions for the test of the 2c diesel generator will be identified as another. example of violation (50-321, 366/86-30-01). This diesel generator test and the concerns expressed indicated poor planning and preparation, inadequate procedure preparation ar.d review and deficiencies in pre-test briefings and trainin Procedure AG-MGR-21-0386N, Evaluation Pre-test Brief Requirements, requires that a pre-test brief be conducted for any safety special procedure. The pre-test brief procedure specifically requires that the brief ensures participating personnel are'made aware of related procedural requirements, current system clearances, and expected system response, parameters and annunciators. -The apparent operator lack of familiarity with the procedures, the caution statements, the expected' parameters, and the non-proceduraly required actions, indicated that the pre-test briefing did not fully meet the intent of the briefing procedur Plant management's initial response to the concerns was basically that they

  ~

could have_ turned the diesel over to the vendor for testing with no procedur This response appears contradictory to the. added emphasis by Georgia Power on procedural- compliance and increased controls over contractor . Design, Design Changes and Modifications (37700) , The inspector conducted an inspection of the implementation of design I changes during Unit 2 outage to assess the adequacy of the following requirements: ! - Procedures have been established to control design changes which

include assurance that a proposed change does not involve an unreviewed

safety question or a change in Technical Specifications as required by

10 CFR 50.5 Procedures and responsibilities for design control have been

established including responsibilities and methods for conducting

safety evaluations.

t

 - Administrative controls for design document control have been I  established for the following:

i l l

Is-- .-- -c._.__ _

m

-
     .
,

 (a) Controlling changes to approved design change documents (b) Controlling or recalling obsolete design change documents such as revised drawings and modification procedures (c) Release distribution of approved design change documents Administrative controls and responsibilities have been established commensurate with the time frame for implementation to assure that design changes will be incorporated into:
 (a) Plant procedures (b) Operator training programs (c) Plant drawings to reflect implemented design changes and modifications
-

Design controls require that implementation will be in accordance with approved procedure Design controls require assigning responsibility for identifying post-modification testing requirements and acceptance criteria in approved test procedures and for evaluation of test result Procedures assign responsibility to delineate the method for reporting design changes to the NRC in accordance with 10 CFR 50.5 Controls require review and approval of temporary modifications in accordance with Section 6 of the Technical Specifications and 10 CFR 50.5 Plant Hatch procedure number 42EN-ENG-001-0S delineates.the requirements for initiation and approval of Design Change Requests (DCRs). Responsibilities for design activities have been assigned, and requirements for performing Nuclear Safety Evaluations have been established in writing. Additional design change program documents provide instructions to Engineering Department personnel on the requirements and methods for review, preparation, implementation, and closing out of DCRs. Based on a review of these procedures, it appears that adequate design controls have been established for the processing of plant modifications.

t

,_ _ _ _ . , _  _ , _ _ . _ _ _ . _ _ _ . _ . _ _ __ - . _ , _ = _ .__ ,

 .
~

.

The inspector conducted interviews with selected members of the onsite Design Change and Modification Group to determine the scope of plant changes being implemented during Unit 2 outage. Consequent to these discussions, the following DCRs were selected for review to verify procedural conformance and technical adequacy of the design change packages: DCR No.: 85-152, Revision 0 System or Component: Moisture Separator MPL No.: 2N38-B001A-D Design Objective: To Prevent Turbine Trip From a Single Spurious Contact Closure DCR No.: 82-46, Revision 1 System or Component: Reactor Manual Control System MPL No.: 2C11 Design Objective: To Protect the Reactor Manual Control System 24 and 28 Y DC Power Supplies From All Combinations of Load and Voltage Variations DCR No.: 78-256, Revision 0 System or Component: LOCA and LOSP Timers MPL No.: R43 Design Objective: Provide More Reliable Timers to be Used for Load Shedding and Sequencing During a LOCA or LOSP Emergency DCR No.: 85-223, Revision 1 System or Component: RWCU Pump MPL No.: 2C31 C001B Design Objective: Replace Existing Union Pump With a Hayward Tyler Sealess Pump DCR No.: 82-171, Revision 0 System or Component: Protective HFA Relays MPL No.: A71, B21, C61, C71, et Design Objective: To Increase the Life of HFA Relays and to Upgrade System Reliability for Those Systems in Which HFA Relays are Installed Pursuant to the review of the above design change implementation packages, the inspector verified that the following activities had been completed:

- Nuclear safety evaluation / nuclear evaluation checklist
- Functional test requirements with minimum acceptance criteria specified
-

Fire protection review

-

Environmental evaluation checklist _ - _ *

.

12 The DCRs were reviewed to assure consistency between the design inputs and the design activitie Completed post modification test results, were available, were reviewed to assure that the design objectives were me One deficiency was identified during this review process, Design Change Request (DCR) 82-171 involved the changeout of 134 HFA relays from circuits which provide protection to components in the residual heat removal (RHR), reactor core isolation cooling (RCIC), high pressure cooling (HPCI), automatic depressurization (ADS), and core spray (CS) system Upon reviewing the logic as depicted on the electrical control schematic for relay 2E41A-K6, the inspector determined that inaccuracies existed in the , narrative description of the logic contained in the Nuclear Safety Evaluation for Procedure 42SP-080886-QG-1-2 Review of the completed test procedure revealed no change in the logic, however, and apparently an error was made in the narrative description writeu The inspector reviewed documentation for 10 HFA replacement relays that were applied in various nuclear safety related system A discrepancy was identiffs d only with relay 2E41A-K6, and is considered an anomaly. Correction of this documentation problem was discussed with the cognizant implementing enginee Based on the review of the design change implementation packages and discussions with the cognizant implementing engineers, it appears that the implementation of DCRs are adequately controlle Licensee management is presently involved in an effort intended to ensure the revision and maintenance of plant drawings that accurately reflect actual plant condition The inspector conducted interviews with engineering personnel from the Southern Company Services and Bechtel Engineering Corporation to determine the status of the As-Built Notice Progra It was determined that licensee management has established a policy regarding the update of plant drawings with time limits as follows:

- There should be no critical drawings in house older than thirty (30)

day Critical drawings should be processed within thirty (30) days of receipt.

'

- Non-critical drawings should be processed within sixty (60) days of receipt, with the exception of non-critical outage ABN drawings which should be processed within 120 day Critical drawings were defined as P& ids, elementary, and one line drawing Aoditionally, the effective date for implementation of the above policy was stated as August 1, 198 . -- _ - . . _- *

.

In discussions with licensee contract personnel, the inspector was informed that critical drawings required to support plant operation have been updated as of July 31, 198 Projected completion date for non-critical drawing updates and the progress being made were also reviewed and discussed by the inspector. The following documents were reviewed in connection w'th this effort:

- Southern Company Services, Inc. , letter to Mr. P. R. Bemis, Georgia '

Power Company, from Mr. William F. Warner, Subject: E. I. Hatch Nuclear Plants - Units 1 and 2, As-Built Monthly Progress Report, dated September 11, 198 I. Hatch Nuclear Plant Monthly Project Status Report, dated August 198 Graphs of Projected and Actual ABN Closeout Schedule for "Bechtel and SCS Critical Drawings with ABNs".

- Graphs for Bechtel and SCS 1986 ABN Program Showing Status of ABNs Received, ABNs Closed, and ABNs Over 120 Day Licensee management has established a Design Change Request (DCR) Closeout Group onsite with responsibilities for determining required actions for closing DCRs assigned to them. The inspector conducted interviews with personnel from this organization to determine the status of the DCR closecut effor Selected DCRs were reviewed to determine the scope of the plant modifications, number of associated ABNs, and number of drawings impacted by the DCR implementatio The following DCRs and associated ABNs were chosen for use in verifying that critical drawings with listed ABNs had been update DCR 84-221, Revision 0, ABN No. 84-502

- DCR 82-004, Revision 2, ABN No. 85-395 ABN No. 83-405
- DCR 80-409, Revision 0, ABN No. 86-457
-

DCR 81-51, Revision 1, ABN No. 82-224

- DCR 78-283, Revision 0, ABN No. 83206 The inspector reviewed 42 drawings consisting of P& ids, electrical elemen-tary diagrams, and electrical interconnection diagrams. A total number of 27 drawings had been revised to incorporate ABNs listed against them. Of the 15 drawings that were not updated, the inspector did not identify any critical drawings as defined by the license From the results of this verification effort, the ABN program appears to be effective in ensuring the update of critical drawings to reflect actual plant configuratio Within this area, no violations or deviations were identifie __ _

-   .
'
.

14

.

8. Radiological Controls of Contract Personnel Subsequent to a tour of the reactor building, the inspectors were exiting the radiological controlled area (RCA) at the C-52 control point and noticed an apparent lack of enforcement for health physics requirements. As the inspectors were frisking in the PCM-1 portal monitors at the main exit area, a contrtcc employee exited the area without frisking in the monitor. The inspectors questioned the three HP technicians present at the exit area to determine if some personnel were exempt from frisking. They stated that all personnel exiting the area must frisk prior to their exit and that they had not noticed the individual bypass the portal monito NRC concern over this issue was expressed at the exit meeting due to the large number of contract personnel presently on site, and the fact the facility is in a major outage with increased levels of contamination present throughout the Unit 2 RC Hatch procedure 62RP-RAD-017-0, Release Surveys for Trash and Materials Leaving Operating Buildings, requires that the Health Physics Technician at C-52 is responsible for ensuring that all persons exiting the control building use the monitors upon exit. The licensee was informed during a conference call on October 21, 1986, that the failure of contract personnel to properly frisk when exiting the RCA is another example of violation (50-366/86-30-01), failure to follow procedures.

i l l l l l l l 1 . .-

      -

}}