ML20150C505

Insp Rept 50-29/88-01 on 880120-0222.No Violations Noted. Major Areas Inspected:Actions on Previous Insp Findings & Operational Safety Verification.Items of Concerns Include Dedicating Commercial Parts for Safety Class Generators

ML20150C505
Person / Time
Site:	Yankee Rowe
Issue date:	03/14/1988
From:	Haverkamp D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20150C502	List: IR 05000029/1988001 ML20150C505
References
50-029-88-01, 50-29-88-1, NUDOCS 8803210019
Download: ML20150C505 (29)

Text

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U.S. NUCLEAR REGULATORY COMMISSION REGION I Report No. 50-29/88-01 Docket No. 50-29 Licensee No. DPR-3 T

Licensee Yankee Atomic Electric Company.

1671 Worcester Road Framingham, Massachusetts 01701 Facility Name: Yankee Nuclear Power Station Inspection At: Rowe,-Massachusetts

! Inspection Conducted: January 20, 1988 - February 22, 1988 ' '

Inspectors: Harold Eichenholz, Senior Resident Inspector Cynthia A. Carpenter ' Resident Inspector Approved By: /M Donald R. Haverkamp, Chief as b/ 3/W/#

Date Reac+.orProjectsSection-No.[3C i i Inspection Summary: Inspection on January 20, 1988 - February 22, 1988 (Report No. 50-29/88-01) ,

Areas Inspected _ Routine onsite regular and backshif t inspection by resident inspectors (161 hours0.00186 days <br />0.0447 hours <br />2.662037e-4 weeks <br />6.12605e-5 months <br />). Areas inspected included licensee action of previous inspection findings, operational safety verification, bimonthly safety system ,

verification, radiological controls, plant events, maintenance observations, t

! surveillance observations, periodic and special reports, on-site review commit- i tee activities, licensee response to selected safety issues, procurement of zine conduit couplings, licensee response to IE Bulletins, and participation in NRR/ licensee meeting.

Results: No violations or deviations were identified. However, two i tems involving inspector concerns or program weaknesses were identified regarding: *

(1) use of special orders in lieu of approved procedures (Section 3), and (2). the acceptability of the licensee's practices associated with dedicating -

L commercial grade parts for the safety class emergency diesel generators (Section 8).

Facility activities during this inspection period were, in general, reflective  ;

of the licensee's strong orientation towards nuclear safety. Areas that con- t tinue to be considered notable licensee strengths involve fire protection and l

housekeeping and maintaining control room annunciators in a black-board status

(Section 3). The licensee continued to strengthen its security program by improvement in management oversight and responsiveness to day-to-day issues and ,

l events (Section 3).

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TABLE OF CONTENTS Page

1. Persons Contacted. . . . . . . . . . . . . . . . . . . . . . . . 1

2. S um.ma ry o f Fa c i l i ty a nd NRC Ac t i v i ti e s . . . . . . . . . . . . . 1

3. Licensee Actions on Previous Inspection Findings (IP 92701)* . . 1 l

4. Operational Safety Verification (IP 71707) . . . . . . . . . . . 2

a. Daily Inspection. .. ................... 2

b. System Alignment Inspection . . .............. 4

c. Biweekly and Other Inspections. . . . . . . . . . . . . . . 4

d. Backshift Inspection. . . ................. 5

5. Bimonthly Safety System Verification (IP 71710). . . . . . . . . 6

6. Radiological Controls (IP 71707) . . . . . . . . . . . . . . . . 6

7. Plant Events (IP 93702) . . . . . ............... 8

a. No. 1 Boiler Feed Pump Motor Shaft Out-of-Round . . . . . . 8

b. Turbine Control Valve Problem . ............. 8

c. Condenser Tube Cleaning . . . . .............. 9

8. Maintenance Observations (IP 62703). . . . . . . . . . . . . . . 9

9. Surveillance Observations (IP 61726) . . . . . . . . . . . . . . 12

10. Periodic and Special Reports (IP 90713). . . . ........ 14

11. On-Site Review Committee Activities (IP 40700) . . . . . . . . . 16

12. Licensee Response To Selected Safety Issues. . . . . . . . . . . 16

a. Low Temperature Overpressure Protection (TI 2500/19). . . . 16

b. Natural Circulation Cooldown (TI 2515/86) . . . . . . . . 18

c. Storage Battery Adequacy Audit (RITI 87-07) . . . . . . . . 19

d. Fitness for Duty (Drug Testing) Information and Report (RITI 88-01). . . . . . . . . . . . . . . . . . . . . . . . 19

13. Procurement of Zinc Conduit Couplings (IP 38701) . ...... 20

14. Licensee Response to IE Bulletins (IP 9%703) . . . . . . . . . . 20

15. Participation in NRR/ Licensee Meeting (IP 94702) . . . . . . . . 21 i

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Table of Contents (Continued)

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16. Unresolved Items . . . . . . . . . . . . . . . . . . . . . . . . 21

17. Open Items . . . . . . . . . . . . . . . . . . . . . . . . . . . 21

18. Management Meetings (IP 30703) . . . . . . . . . . . . . . . . . 21 Attachment 1 - NRC:RI Temporary Instruction 87-07, Attachment 2

• The NRC Inspection Manual inspection procedure (IP) or temporary instruction (TI) or the Region I temporary instruction (RITI) that was used as inspection guidance is listed for each applicable report section.

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r DETAILS l t

l '. Persons Contacted.

- Yankee Nuclear Power Station-(YNPS)

N. St. Laurent, Plant Superintendent B. Drawbridge, Assistant Plant Superintendent T. Henderson, Technical Director I

Yankee Atomic Electric Company (YAEC)

J. DeVincentf s, Vice President and Acting Manager of Operations D. Maidrand, Assistant Project Manager G. Papanic, Licensing Engineer j 4

The inspector also interviewed other licensee employees during the inspec-tion, including members of the operations, radiation protection, chem- ,

istry, instrument and control, maintenance, reactor engineering, security, training, technical services and general office staffs. '

2. Summary of Facility and NRC Activities At the start of the inspection period on January 20, 1988, the plant was -

at 75% of rated power during repair of the No. I boiler feed pump (EFP). -

The No. 1 BFP motor shaf t was found to be out-of-round. The BFP was '

returned to service and the plant was at 100% of rated power on

! January 23, 1988. The plant remained at full power until February 20, 1988 ,

l when the licensee commencec', a planned load reduction to approximately 50% '

of rated power in order to perform maintenance, particularly main conden-

ser tube cleaning. On February 22, 1988, as the inspection period ended, t

the plant was returning to full power operations.

Two NRC Region I (NRC:RI) health physics specialist inspectors completed an inspection in the area of radiological controls during the period of l February 8-12, 1988 (Inspection Report No. 50-29/88-04). Also during the l period of February 8-12, 1988 an NRC:RI physical security specialist com-pleted reviews of the licensee's physical security program (Inspection Report No. 50-29/88-03). During the period of January 24-29, 1988, the

, Yankee Resident Inspector substituted for the Resident Inspector at r Seabrook Station, Unit 1, Seabrook, New Hampshire.  !

3. Licensee Action on Previous Inspection Findings t t

! (Closed) Inspector follow item 50-29/85-24-02: Review design details and [

i circumstances that necessitate an expansion tank to allow pressure regu-i lating valve SI-PR-53 to function properly. The inspector reviewed the i licensee's justification fo" +he addition of an expansion tank for SI-PR-53. This expansion tar p 'ovides additional surge capacity to limit t

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. 2 the pressure drift of one of the three regulator valves. Since the regulators are set the same, it appears that the expansion tank is not directly related to a problem with the specific regulator but most likely to the way the entire system is configured. This resulted in SI-PR-53 being a little more difficult to control without the expansion tank. This item is closed.

(Closed) Inspector follow item 50-29/85-15-01: Review licensee corrective actions associated with ASCO pressure switch setpoint drift. The NRC Team Inspection 50-29/88-02 reviewed licensee corrective actions on this item.

In addition, licensee corrective actions were reviewed in Section 8 of NRC Inspection Report 50-29/87-16. This item is considered closed since resolution of any remaining NRC concerns about the ASCO pressure switch failures will be resolved in NRC Inspection Report 50-29/88-02.

4 Operational Safety Verification

a. Daily Inspection During routine facility tours, the inspector checked the following items: shift manning, access control, adherence to procedures and limiting conditions for operations (LCOs), instrumentation, recorder traces, protective systems, control rod positions, containment tem-perature and pressure, control room annunciators, radiation monitors, radiation monitoring, emergency power source operability, control room and shift supervisor log, tagout log and operating orders. Based upon a review of licensee activities in this area the inspector noted the following:

(1) As a result of reviewing the Rowe Station Log Sheet No. 2 on January 23, 1967, the inspector noted that the control room operators had identified that the pressurizer narrow range pressure pen on the MC-PLR-101 recorder was acting erratic.

This condition was also logged in the off-normal equipment list-ing of the shift turnover log, APF-2002.1. The inspector ver-ified that the control room operators had issued maintenance request (MR)88-008 to initiate maintenance activity to provide a mechanism to resolve this item.

(2) During a tour of the control room on January 28, 1988, the inspector noted a January 27, 1988 6:00 p.m. control room log entry that indicated that the operators had blown down all the steam generator pressure sensing lines. The inspector question the senior control room (SCRO) as to the nature of the activity, and was informed that this action was being taken at the request of the plant operations manager in response to problems encoun-tered with the sensing lines that are caused by cold weather.

When questioned as to the manner in which this activity is con-troled, the SCR0 indicated that special order No. 88-08 issued on January 14, 1988 was used.

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. 3 The inspector reviewed that special order 'and noted that it pro-vided a sketch and specific instructions in the manner in which the pressure sensing lines were to be blown down. Administra-tive procedure AP-2006, Rev. 6, Special Orders specifies the administrative controls for the development and issuance of management instructions 'that' encompass orders pertaining to special operations, housekeeping, data taking, etc., and revis-ions to procedures and/or'surveillances that Operations Depart-ment personnel should be aware of. During the prior SALP period (SALP Report 50-29/85-98), the NRC specified the need for the -

licensee to insure that the special orders were not being used in lieu of-approved procedures. In addition, that SALP ' Report specified the need for the -licensee to insure that procedures are developed for all planned operations. Within the current SALP period NRC Inspection Reports 50-29/86-19 (Section 4) and.

50-29/87-04 documented the need for licensee management atten-tion to address these NRC concerns. The issuance of Special Order No. 88-08, has again resulted in the inspector questioning the licensee's practices of not developing an operational pro-cedure to cover the specified task. This matter remains an open item (50-29/88-01-01), pending a review of licensee actions to strengthen operations management awareness of the appropriate use of special orders.

(3) Throughout the inspection period, the inspector noted an excel-lent level of performance of the licensee in maintaining the control room annunciators in as close to a "black-board" status as possible.

(4) On February 19, 1988 the inspector noted that shutdown rod groups D, B and A had been withdrawn from a position of 89 2/G 1 to 89 5/8 in accordance with AP-7104, Rev. 69, Core Operational Limits. The licensee indicated that the reason for the with-drawal of the shutdov:n rod groups was that the burnup on the core was now greater than 7500 MWD /MTV and as required by AP-7104, the shutdown rod groups were withdrawn to 89 5/8. This action distributes wear by the guide blocks on the control rods and the 89 5/8 is the maximum withdrawal administratively. The licensee's actions were confirmed to be in accordance with pro-

! ceduro AP-7104 and the procedure and licensee action was in accordance with TS requirements. The inspector had no further questions.

No violations were identified during resident inspector daily inspections.

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b. System Alignment Inspection Operating confirmation was made of selected piping system trains.

Accessible valve positions and status were examined. Power supply and breaker alignments were checked. Visual inspections of major components were performed. Operability of instruments essential to system performance was assessed. The following systems were checked during plant tours and control room panel status observations:

Emergency diesel generator units Low pressure and high pressure injection systems Charging System Motor driven emergency feedwater pumps No violations were identified during system alignment inspections,

c. Biweekly and Other Inspections (1) General Facility Observation During plant tours, the inspector observed shift turnovers, com-pared boric acid tank sample analyses and tank levels to Tech-nical Specifications requirements, and reviewed the use of radiation work permits and radiation protection procedures.

Area radiation levels and air monitor use and operational status were reviewed. Verification of tagouts indicated the action was properly conducted.

No violations were identified during resident inspector plant tours.

(2) Fire Protection and Housekeeping Also during plant tours, the inspector observed licensee house-keeping and fire protection practices. A strong commitment to proper housekeeping conditions and practices by the plant staff is routinely observed by the inspector. Performance in this area continues to be viewed as a licensee strength.

No violations were identified.

(3) Observations of Physical Security Selected aspects of plant security were reviewed during regular and backshif t hours to verify that controls were in accordance with the security plan and approved procedures. Based on a review of licensee activities in this area, the inspector noted the following:

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On February 18,1988 at 4:18.p.m. , a security 'of ficer was heard to be in a verbal altercation with another security l

officer. Upon hearing the loud argument, the on- duty security shif t supervisor had .the security' officer who had become very agitated come into his office, the security

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of ficer was disarmed within two minutes, . his key card invaiidated and he was suspended from his duties until further investigation by the security contractor and subse-quent licensee review. The inspector was informed of this event in a timely manner. The inspector noted that' the.

licensee and security contractor took appropriate actions to diffuse the situation,- re verify security measures and investigate the situation in a timely and comprehensive manner. On February 19, 1988, NRC:RI security management was appraised of the event by the inspector and the licen-see's Security Manager. The NRC had no further questions of the licensee on this matter.

As noted in Inspection Report 50-29/87-16, a problem existed with access control equipment at the gatehouse that involved alarm thresholds that were overly sensitive. Dur-ing this inspection period, the licensee had completed its evaluatica of equipment performance and calibration requirements, and had taken appropriate corrective actions

.that were responsive to the inspector's concerns. This issue is considered resolved. As a result of the improve-ments in licensee oversight in the area of security equip-ment maintenance, the resolution of security system and equipment malfunctions wn :, provided in a timely manner.

Improvements in station management oversight of security activities continued to be observed by the inspector. The January 31, 1988 status listing of the security program improvement initiative issued by the Manager, Administra-tive Services was reviewed by the inspector. No unaccept-able conditions were identified.

No violations were identified during observations of physical security.

d. Backshift Inspection The inspector conducted backshift, weekend or holiday inspections on January 21, 23, and 28, and February 2, 7, 8, and 19. Operators and shift supervisors were attentive and responded appropriately to annunciators and plant conditions. No violations were identified during backshift inspections.

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5. Bi-monthly Safety System Verification The inspector independently verified the operability of a selected engi-neered safety features (ESF) system by performing a complete walkdown of the accessible portions of the system to:

confirm that the licensee's system lineup procedures match plant drawings and the as-built configurations; identify equipment conditions and items that might degrade performance; verify appropriate levels of cleanliness were being maintained; verify technical specification requirements are adhered to; verify instrumentation lineup and calibratian; and verify proper valve position, availability for function and position indication.

The low pressure safety injectier. accumulator system was reviewed. The inspector noted that ECCS accumulator vent line drain valve SI-V-42 on top of the accumulator appeared to have boron precipitation around the valve cap and body; this was brought to the licensee's attention. The licensee verified that it was boron precipitation, performed a swipe and found the valve cap to be contaminated. The valve was subsequently cleaned and it was determined that the boron was due to a packing leak. An MR has been issued.

The inspector also reviewed the circumstances that necessitated an expan-sion tank to allow pressure regulating valve SI-PR-53 to function properly (inspector follow item 85-24-02). See Section 3 of this report for reso-lution of this item.

No violations were identified during the bi-monthly safety system verification.

6. Radiological Controls Radiological controls were observed on a routine basis during the report-ing period. Standard industry radiological work practices and conformance to radiological control procedures and 10 CFR Part 20 requirements were observed. Independent surveys of radiological boundaries and random surveys of nonradiological areas throughout the facility were taken by the inspector.

7 On several occasions the inspector observed the licensee taking primary. -

and secondary samples. Primary samples are taken in the primary sample sink under an exhaust hood; the exhaust hood has a glass front window which is rolled up or down to protect personnel from possible airborne contamination. The area in front of the primary sample sink has a radio-active materials rope across it, step-off pad and magenta and yellow tape ,

across the floor to mark the area where the radioactive materials area begins. The chemistry technician was observed, although in coveralls, reaching from outside the roped area to inside the roped area to move an item and, in one case, with one pair of rubber gloves on and after hand-ling a primary sample, reaching out of the marked area behind him to obtain additional paper towels and rubber gloves.

These occurrences do not illustrate good radiologiczi work practices. When-a health physics technician was asked about these practices, he indicated that reaching across a radiological boundary is not standard practice. It should be noted that on the second occasion when the inspector observed the drawing of primary and secondary samples, being performed by a differ-ent chemistry technician, good radiological work practices were observed with the exception of reaching over the radiological boundary. The glass front on the primary sample sink was lif ted just enough that he could reach his hands under and obtain the samples; gloves and paper towels were staged inside the hood for use, therefore ensuring that the technician did not have to reach outside the boundary to obtain these items.

During a routine tour of the plant on February 8, 1988, the inspector noted that the licensee had eliminated a high radiation exclusion area (HREA) in the primary auxiliary building (PAB) corridor. Access to the low pressure surge tank (LPST) cooling and the shutdown cooling (SDC) cubicle area, located in the PAB corridor, was no longer locked and posted as a HREA.

The licensee was able to take this action as a result of implementing temporary change requests (TCRs)87-375, 87-376 and 88-443, which installed permanent lead shielding on SDC valves and piping, and the LPST cooler.

The inspector reviewed licensee documentation, determined that proper engineering evaluations were provided by the Yankee project jroup of the Yankee Nuclear Services Division, and that installation of the TCRs was performed in accordance with procedure AP-0018, Rev. 14, Temporary Change Control. The inspector had no further questions of the licensee on this item.

No violations or radiation safety concerns were identified during radio-logical controls observations.

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. 8 7 ', Plant Events

a. No.1 Boiler Feed Pump Motor Shaf t Out-of-Round At the beginning of _this inspection period, the plant was at-75% of rated power. This was due to the No. 1 boiler. feed pump (BFP) motor shaft being found ' out-of-round, causing the ' oil ' seals to not 'be successful in sealing the shaf t. The motor shaft was removed and sent to the manufacturer for repair. The plant remained at' 75% of rated power during repair of the_ BFP motor shaf t. On . January 22, 1988, the BFP was returned to service and the plant started a load increase from 75% of rated power; at 2% per hour. The plant was returned to full power operation at 5:00 a.m. on January 23, 1988.

b. Turbine Control Valve Problem At 12:00 a.m. on February 20, 1988 the licensee commenced a control-led load reduction to 77% of rated power to conduct turbine valve testing in accordance with OP-4225, Rev. 17, Turbine Throttle and Control Valve Surveillance Test. This high risk surveillance in-volves opening tnd closing of throttle and control valves in sym-metry. Plant load is reduced until No. 3 and No. 4 control valves (CV) are closed. The No. 4 CV is then locked closed using the test device. Then, using the test device, the No. 1 CV is closed fully by-opening NO. 3 CV from the control board. Using the test device, the No. 1 CV is opened fully while closing the No. 3 CV from the control board. At this time, the No. 3 CV is locked closed with the test device. Next, the No. 4 CV is opened from the control board while the No. 2 CV is slowly closed using the test device. Finally, the No. 2 CV is opened using the test device while closing No. 4 CV from the control board. N o .. 3 CV is unlocked, plant load is picked up until all CVs are fully open and/or plant load is at the maximum allowed power output.

At 4:30 a.m. af ter closing the No. 4 CV and while reducing governor oil pressure to close the No. 3 CV, when the No. 3 CV was half way closed, tha valve unexpectantly went almost entirely closed. How-ever, it did not close completely. The licensee continued to lower oil pressure in attempting to close the No. 3 CV completely. Also, l they attempted to stroke the valve manually by working on its pilot servomotor. This also failed to completely close the valve.

l l As the licensee lowered the governor oil pressure to fully close No.

l 3 CV, the No. 2 CV began to close. This caused the load to drop off rapidly and cold leg temperaSre (Tc) to rise to greater than 520 degree F. The TS 3.2.4 requires the highest operating loop cold leg l temperature to be maintained less than or equal to 520 degree F. If this limit is exceeded, the TS requires the parameter to be restored within its limit within two hours or reduce thermal power. After the

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I". 9 No. '2 CV started to close, the oil pressure was. brought back up to open No. 2 CV fully, and to attempt to open the No. 3 CV. Re-opening of the No. 3 CV ~was unsuccessful. Next, the No._1 CV was closed manually in the expectation that this would open the No. 3 CV; it did not open. The licensee, considered the problem that the No. 3 CV would not' reopen to have been caused by the excessive differential pressure across the No. 3 valve.

Consequently, the No. 1 and No. 2 CV's were opened, lowering the differential pressure across the No. 3 CV. . The oil pressure was brought back up on the valves and the No. 3 CV was manually stroked open and the No. 1 CV closed by use of the test device. Upon com-pletion of this troubleshooting the No.1 and 4 CV's were closed, No.

2 and 3 CV's were open and Tc had been restored to within TS require-ments at 5:15 a.m.

The licensee has not been able to determined the root cause of prob-lem with the No. 3 CV. However, new servomotors have been ordered for the No. 3 ard No. 4 CV's and the licensee intends to replace these at the next outage.

c. Condenser Tube Cleaning On February 20, 1988, upon completion of the throttle and control valve portion of OP-4225 and partial stroking of the main steam non-return valves, the licensee continued the controlled load reduction to approximately 50% of rated power in order to perform 0P-2702, Rev.

9, condenser water box tube cleaning and/or leak check. Only one main. condenser tube required plugging, due to several scrapers becom-ing lodged inside the tube. Also, during tube cleaning, the casing on one of the two condenser tube cleaning pumps broke, leaving enly one pump for the condenser tube cleaning. This extended the length of the load reduction by approximately 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br />.

No violations or deviations were identified in the review of plant events.

8. Maintenance Observations The inspector observed and reviewed maintenance and problem investigation l

activities to verify compliance with regulations, administrative and main-tenance procedures, codes and standards, proper QA/QC involvement, safety tag use, equipment alignment, jumper use, personnel qualification, radio-I logical controls for worker protection, fire protection, retest require-l ments and reportability per Technical Specifications. The following l activities were included:

i Maintenance request (MR)88-252, No. 3 Bailey feedwater regulating valve inlet flange leak l

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.MR 88-297, No. 1, non-return valve (NRV) train A 10% test circuit did not operate MR 88-298, No. 2 NRV train A 10% test circuit did not operate MR 88-299, No. 4'NRV train A 10% test circuit-did not operate MR 87-1590, No. 2 emergency diesel generator (EDG), defective hose fittings on fuel oil return line and engine oil pressure switch MR 87-1706, No.1 EDG, replace fuel and oil hoses and fittings where needed MR 87-1707, No. 3 EDG, replace fuel and oil hoses and fittings where needed MR 87-1751, No. 2 EDG failed to stop from safety injection panel and local control panel MR 87-1827, No. 2 EDG replacement of solenoid dump line hose Based upon a review of licensee activities in this area the inspector noted the following:

The inspector reviewed the licensee's repair activities associated with MRs88-297, 88-298, and 88-299, which covered the corrective maintenance for the Nos. 1, 2, and 4 NRV train A 10% test circuits, respectively. The inspector's review is documented in Section 9 of this report.

As a result of a facility tour conducted by Mr. William Russell (NRC Region I Regional Administrator) on September 16, 1987, the licensee reviewed the condition of fuei and oil hoses on the Nos. 1, 2, and 3 l EDGs. The MRs 87-1590, 87-1706, and 87-1707 were issued by the licensee to conduct the review and effect corrective maintenance, i Diesel repair services are performed by either the plant maintenance department or the licensee's diesel repair service, Power Product.

A blanket purchase order No. QA40150 was issued on Cs ter 26, 1986 to Power Product for materials and services for the YNPS's three EDGs, and covers the year 1987. A provision of this purchase order stipu-lates that the service / items are to be performed in accordance with the vendor's quality assurance program which has been approved by YAEC. In accordance with the licensee's material. and service pur-chase procedure AP-0211, Rev.14, the purchase order was issued with the understanding that the vendor was on the approved vendors list (AVL). However, on January 21, 1987 the quality assurance department (QAD) removed the vendor from the AVL because of a 1ack of a docu-mented quality assurance program. A quality assurance department

11 memorandum, QAD 87-149/4-5, issued on February 13, 1987 specified, in part, that diesel repair services performed at Power Product or at subtier vendor facilities will require YNSD QAD source surveillance to assure compliance with procurement documents, and that diesel spare parts provided by Power Product should be purchased as commer-cial grade whenever possible. This memorandum contends that although the vendor will no longer be considered as an approved vendor, imple-mentation of the above stated actions should provide requisite assur-ance of the quality of items and services purchased from the vendor.

The inspector noted the purchase order also specified that a certif-icate of compliance shall be provided for the material supplied to this purchase order. All certificates of compliance supplied for the purchase order stipulated that the material complies with the terms and requirements of the subject purchase order and are approved spare parts for the licensee's EDGs. It appears that the licensee never modified the purchase order to indicate that the vendor was no longer on the AVL.

Replacement of hoses and fittings where needed was performed on October 14, 1987 on the No. 1 EDG by Power Product with the assist-ance of maintenance department personnel. The MR 87-1706 was used, with procedure OP-5350, Rev. 9, inspection and/or maintenance of the diesel generator system used as the controlling maintenance proced-ure. Appropriate retest was performed and the EDG was returred to service. Similarly, No. 2 EDG had its hoses and fittings inspected and repaired on October 21, 1987. During the performance of post-maintenance testing, the EDG would not shut down from either the local or remote control panels. Because the engine oil was changed out as part of the maintenance activity, the maintenance department requested that the EDG be run for two nours under load. The MR 87-1751 was issued and closed on October 21, 1987 as a result of this identified problem. The EDG had satisfactorily shut down at the end of the two-hour run. As a result of subsequent observations that the EDG was shuting down in a sluggish manner, MR 87-1827 was issued on November 2, 1987 to replace the solenoid dump line hose. This activ-ity was conducted on November 3, 1987 and resulted in determining that the new high pressure flexible hose installed on October 21, 1987 hao contained a restriction, preventing the oil to drain fast enough to trip the fuel rack. The restriction occurred during the assembly of the fittings to the hose by a Power Product technician.

The hose was replaced and the EDG was satisfactorily tested.

The QAD initiated an investigation into the f ailure on Oaober 21, 1987 for the No. 2 EDG to shut down. This investigation was docu-mented in memorandum QAD 88-103/7.1.9 and issued to the plant main-tenance manager on February 10, 1988. The QAD investigation con-cluded that Power Product does provide adequate controls to assure

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that only factory authorized parts or equivalent replacement parts are supplied for the EDGs. Additionally, it was determined that YNPS personnel had failed to implement procedure AP-0212,' Control of Purchase Material, Equipment and Service, in the failure to perform receipt inspection prior to the performance of the work. Receipt inspection was conducted af ter the work had been completed on parts turned into the stock room. On December 29, 1987, the plant Quality Assurance Supervisor initiated a nonconformance report (No. 87-42) to resolve this deficiency and provide corrective action. As of the completion of this inspection period, this nonconformance report has not received the attention of either the plant operations review committee or licensee management.

The manner in which the licensee dedicates the commercial grade components it receives from Power Product for use in the safety class EDGs remains an unresolved item (50-29/88-01-02), pending review of licensee actions during a subsequent NRC inspection.

9. Surveillance Observations l

The inspector observed tests or parts of tests to assess performance in accordance with approved procedures and LCOs, test results (if completed),

removal and restoration of equipment, and deficiency review and resolution.

The following tests were reviewed.

OP-4214, Rev. 12, Chemical Shutdown System Operability Check OP-7104, Rev. 69, Core Operational Limits OP-4260, Rev. 3, Turbine Control Valve Exercise OP-4211, Rev. 21, Emergency Feedwater System Operability Test OP-4601, Rev. 20, Nuclear Instrumentation Channels Functional Test l

OP-9406, Rev. 7, Primary Plant Liquid Sample Points OP-9001, Rev. 8, Primary Chemistry Test Frequencies and Specifications OP-4261, Rev. 5, Main Steam Non-Return Valve (NRV) Operability Test OP-4265, Rev. 17, Turbine Throttle and Control Valve Surveillance .

Test OP-4201, Rev. 13, Power Range Channel Calibration Heat Balance l

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. 13 Based upon a . review of licensee activities in this area, the inspector noted the following:

On January 20, 1988 the operations department implemented a ' monthly

. flow test of the boric acid mix tank (BAMT) . suction line in accord-ance with Attachment B to procedure OP-4214. This . procedure requires that demineralized water will be backflushed through valve CS-MOV-529 and then enters the BAMT. Operators are expected to note a slight increase in the BA'4T level. Although the procedure requires . the operators to notify the chemistry department that this procedure will be performed, and to sample the BAMT at the appropriate time, the procedure acceptance criteria did not include satisfying the condi-tions of TS 3.1.2.11 requirements that 1) the concentration of the boric acid solution be between 1294 and 12.5?;, and 2) that the minimum

. solution temperature will be 150 degree F. In addition, the proced-ure does not require the operator to enter the TS action statement at the beginning of the surveillance test, even though the possibility exists for the test to create a loss of operability.

As a result of the performance of this surveillance test, the chem-istry ' department notified the control room at 1:30 p.m. that the concentration was below the TS limit. The control room operators acknowledged in the control room log that the action vtatement of TS 3.1.2.11 allowed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> period in which this condition was accept-able for continued plant operations, and initiated corrective measures to return the BAMT to operational status. The control room logged the completion of procedure OP-4214 at 3:00 p.m. on January

!. 20, 1988 and the return of the BAMT to normal operation and exiting action statement at 3:10 p.m. the same day.

Although the inspector intended to hold discussions with operations department personnel about the naed to address the aforementioned concerns, independently the operations department revised the proced-l ure. On February 4, 1988, Revision 13 to OP-4214 was issued as a general update of the procedure. This revision was reviewed by the inspector and was determined to fully resolve the inspector's con-cerns. The inspector had no further questions of the licensee on this matter.

During the performance of OP-4261 on February 20, 1988 to partially (10?e) stroke each non-return valve while at full power the Train A 10?; test circuit did not actuate on the No. 1, 2 and 4 NRV's. Each main steam NRV is required to be demonstrated operable in Modes 1-3, by cycling each valve through at least 1094 of full travel. During this test, each NRV is tested on both Train "A" and Train "B". The valve is partially stroked by use of a separate test circuit, one for

14 each train. By energizing a solenoid in the NRV control test circuit of either train, the NRV will begin to clase until the valve has moved 10% (90% open) in the closed direction. At this time, the solenoid is de-energized, closing the main dump valve and starting the. hydraulic oil pump which returns the NRV to the full ' open position. Operation of this test circuit is completely independent-of the main dump solenoid, ensuring operability of the circuit and NRV.

There are two independent trains for operating the NRV. Both trains.

must function to open the valve, but only one train is required to function to close the valve. During this surveillance, the Train "A" solenoids in the test circuits for NRV No. 1, 2 and 4 failed to-actuate when energized. The Train "B" solenoid in the test circuit was immediately tested and in all cases actuated properly. The licensee issued MRs88-297, 88-298, and 88-299 for the Nos. 1, 2, and 4 NRV train A 10% test solenoid circuits, respectively, to control the maintenance activity. The solenoid internal poppets on the failed Train A solenoids were nudged and immediately actuated properly. The Train A solenoids were tested three more times and actuated each time when energized. Troubleshooting indicated the problem was with the actual test solenoid, not in the control circuit.

The solenoids in question were manufactured by KEANE. The KEANE solenoids have a history of problems and the licensee plans to re-place these solenoids at the next outage with a different manufac-turer's solenoids. The inspector reviewed the licensee's actions with respect to this problem and ensured that TS requirements had been met. The licensee took prompt corrective action to determine the cause of the problem, adequately ensured themselves that TS requirements were met and that the failure of the test solenoids would not prevent automatic closure of the NRV's should the valves receive a signal to close. The inspector had no further questions.

No violations were identified during surveillance observations.

10. Periodic and Special Reports Periodic and special reports submitted to the NRC pursuant to Technical Specification 6.9 were reviewed. The review ascertained: inclusion of information required by the NRC including test results and/or supporting information; consistency with design predictions and performance specifi-cations; adequacy of planned corrective action for resolution of problems; determination whether any inf 4 tion should be classified as an abnormal occurance; and validity of i . ed information. The following periodic reports were reviewed.

F-.

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. 15 Monthly statistical reports submitted per TS 6.9.3 for the months of August 1987.through January 1988.

Core XIX Startup Program for the YNPS submitted per TS 6.9.1. dated September 30, 1987.

Special report submitted per TS 4.4.3.1 and 6.9.6 containing the results of the~ inservice inspection -(ISI) examination of ASME code class 1, 2 and 3 components dated September 18, 1987.

Reactor Containment Building Integrated Leak Rate Test dated.

December 16, 1987.

t l During the review of the Core XIX Startup Program analysis for compliance l to TS 6.9.1, the inspector noted that the TS requires the report to address each of the tests identified in the FHSR. This requirement should

! be changed to recognize that the FHSR has been replaced with the Final i

Safety Analysis Report (FSAR). Likewise, the inspector verified that the Core XIX Startup Program complies with the . licensee's startup program commitments as stated in the YNPS Cycle 19 Core Performance Analysis report, attached to a licensee letter to NRC:NRR dated March 11, 1987.

l During review of the intervice inspection examination report for compli-l ance to the requirements of TS 4.4.9.1 the inspector observed a discrep-I' ancy in TS 3.4.3 requirements. The TS 3.4.9 requires the structural l integrity of ASME Code Class 1, 2 and 3 components to be maintained in F accoraance with TS 4.4.9.1, 4.4.9.2 and 4.4.9.3. The TS surveillance requirements 4.4.9.2 and 4.4.9.3 were requested by the licensee to be deleted by Proposed Change (PC) No. 186 dated July 19, 1985. The NRC <

issued TS amendment No. 91, deleting the requirements for the inspection l l of the involved components per TS 4.4.9.2 and 4.4.9.3 on January 15, 1987. l Although TS 4.4.9.2 and 4.4.9.3 were deleted by amendment No. 91, the TS 3,4.9 requiremant still refers to those two surveillances. Additionally, amendment No.104 issued on May 7, 1987 which changed TS 4.4.9.1 did not j

! reflect that TS 4.4.9.2 and 4.4.9.3 had been deleted from the TS l requirements.

During review of the Reactor Containment Building Integrated Leak Rate

Test Report the inspector observed that the report was issued on

December 16, 1987. The Type A Test was completed in May 1987. Also enclosed was the licensee's separate accompanying Summary Report of Type B and Type C results and results of the Type A Test Performed May 1987.

Those reports were submitted to the NRC in accordance with the require-ments of 10 CFR 50, Appendix J, Paragraph V.B including requirements that l those reports be submitted to the NRC approximately three months after the conduct of each test. Additionally, Technical Specification 4.6.1.6 requires that upon completion of determining the structural integrity of i

b

, 16 the containment vessel for each Type A containment leakage rate test, a detailed report shall be submitted to the NRC pursuant to Specification 6.9.6 within 90 days af ter completion of the surveillance. The inspector also noted that licensee procedure AP-0007, Rev. 6, Reports and Postings, requires the Reactor Containment Building Integrated leak rate test, per 10 CFR 50, Appendix J, Para V.B. to be submitted to the commission within three months.

Discussions with the station personnel indicated that they were aware that the reports would be late; the licensee was waiting on analyses from Yankee Nuclear Services Division. However, the NRC was not notified that the required reports would be late, nor was an extension requested. It is noted that although this is a violation of NRC requirements, the licen-see's technical specifications and the licensee's procedure AP-0007, this appears to be an isolated case. Other reports required per the TS have been timely. The inspector also verified that previous Reactor Contain-ment Building Integrated Leak Rate reports have been issued in a timely manner. Likewise, the inspector was informed and aware that the May 1987 Type A test failed to meet the acceptance criterion prescribed f n Appendix J III.A.5(b). The details of the inspector's review of this event are documented in IR 50-29/87-06. The licensee submitted an LER to the NRC on July 11,1987 af ter the discovery of an excessive leakage rate during the testing conducted in F. ough July 1987. Although the inspector con-siders this occurren le a violation, due to the above mitigating circumstances, a Not, , Olation will not be issued.

11. On-Site Review Committee Activities On January 28, 1987 the inspector observed the meeting of the YNPS on-site I

review committee to ascertain that the provisions of TS 6.5. were met.

No violations were identified in the review of this area.

12. Licensee Response To Selected Safety Issues

! As requested by the NRC's Of fice of Nuclear Reactor Regulation (NRC:NRR)

( and NRC:RI, the inspector reviewed the licensee's response to selected l safety issues. Four safety issues were reviewed.

a. Low Temperature Overpressure (LTOP) Protection l NRC Inspection Manual Temporary Instruction 2500/19 required the inspector to conduct a review of licensee activities. The primary I purpose of this review was to verify that the licensee has an effec-tive mitigation system for LTOP transient conditions. The LTOP system provides indications, alarms and automatic functions to prevent overpressurization of the main coolant system at low temperature.

-0,

. 17 The items verified in regard to the LTOP system included design, administrative controls ~and procedures, training, equipment modifi-cations and. surveillance. The inspector reviewed licensee documen-tation forwarded to the NRC and the NRC's responses and subsequent safety analyses conducted by'the NRC on the LTOP system. In a letter-dated August 11, 1976, the NRC requested the licensee to evaluate their system design to determine susceptibility to overpressurization events. In 1976 to 1978, the licensee submitted to the NRC plant specific analyses and proposed changes to the TS in support of the proposed reactor vessel LTOP system for the YNPS. The NRC reviewed the information submitted in support of the proposed LTOP system and in the safety evaluation performed by NRC:NRR supporting TS Amendment No. 59, dated September 14, 1979, the NRC staff concluded that the administrative controls and hardware changes made by the licensee provided adequate protection for the plant from pressure transients at low temperatures. The staff also concluded that the electrical, instrumentation and control aspects of the overall LTOP system design and the related TS changes were acceptable on the basis that they met the criteria relative to (1) operator action, (2) single failure, (3) system testability and (4) seismic category 1 and IEEE Std 279-1971. Amendment 59 to the TS was issued for assurance of proper operation of the LTOP system.

In 1983, if censee review of the design basis for the reactor vessel LTOP indicated that due to shif ts in the Appendix "G" curve , there were events as postulated in the original design basis that could potentially exceed the pressure restrictions of the revised curve for the present cycle (this was discussed in LER 50-29/83-13). In order to maintain the similar levels of protection as demonstrated in the original design basis, the licer.see proposed several administrative /

procedural cnanges to be immediately incorporated and stated that they would evaluate the need for TS revisions, LTOP system modifica-tions and further evaluations. In 1984, the NRC performed a safety evaluation related to the "Appendix G" curve shift affect on LTOP, as requested by the licensee. The staff concluded that the licensee's administrative measures to prevent recurrence of the event were ade-quate and met the requirements specified in 10 CFR 50 Appendix "G" and TS. The staff recommended and the licensee concurred that variable setpoints be used for the power operated relief valve (PORV) during heatup and cooldown as opposed to a fixed setpoint.

Concurrent with the evaluation of the LTOP system design basis, the jicensee, per its letter FYR 85-76 dated July 10, 1985, monitored the revision of Regulatory Guide 1.99 to evaluate the impact that the revised document might have on ' heir modifications to the LTOP sys-tem. The licensee reviewed recr t draf ts of Regulatory Guide 1.99 with the results showing that substantial benefits would be gained in

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. 18 its use for adjusting the Appendix G curve, eliminating the need to make system modifications. The licensee suspenc' J activities on the planned modifications until the draf t regulatory guide is finalized and published. As of this inspection report, the revision to the regulatory guide has not been issued. These safety evaluations verified the design of. the LTOP to prevent reactor vessel overpress-ure conditions.

The inspector also reviewed licensee admid strative controls and pro-cedures for plant heatup and cooldown and verified that the licen-see's administrative measures complied with the licensee's commit-ments that they were being followed during plant heatups and cool-downs and th.c they were properly incorporated into the licensee's procedures. The inspector verified that licensed operators have received training on the LTOP system, that alarm setpoint and'proced-ural controls exist to alert the operators to "arm" the LTOP system and that periodic surveillance of the PORV setpoint is performed in accordance with TS requirements. The inspector selected procedures that require the operator to take action to de-energize equipment during plant cooldown and verified the procedures had been followed.

The emergency core cooling system (ECCS) is removed from service dur-ing plant cooldown in order to eliminate potential for reactor vessel low temperature overpressurization by the inadvertent operation of the ECCS.

LER 50-29/83-13 was reviewed in Inspection Reports 50-29/83-06, 50-29/85-04, 50-29/85-14 and closed in 50-29/85-24.

This temporary instruction is closed, t

b. Natural Circulation Cooldown l The NRC Inspection Manual Temporary Instruction 2515/86 required the

! inspector to conduct a review of licensee activities. The purpose i of this review was to verify that the licensee has implemented a pro-gram for the control of natural circulation (NC) cooldown in accord-ance with its commitments to Generic Letter (GL) iio. 81-21.

l l The inspector reviewed the licensee's commitments in response to GL No. 81-21 and verified continuing implementation of those commit-ments. The inspector likewise verified thtt the licensee's training program included classroom and simulator coverage of procedures on natural circulation cooldown, including recognizing indications that natural circulation cooldown is in progress, symptoms of upper head l voiding and how to avoid upper head voiding. The inspector also j verified that the licensee had an emergency procedure available for the prevention and mitigation of upper head voiding.

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The NRC reviewed the licensee's submittal in response to GL 81-21 and concluded from information provided in its letter dated October' 25, 1983 that there is reasonable assurance that steam formation at the upper head of the reactor vessel during natural circulation cooldown will not occur. 'The staff found that upon acceptable implementation of the NRC-required upgraded Westinghouse Owners Group Emergency Response Guidelines, the licensee's procedures will be adequate to perform a safe natural circulation . cooldown. As stated above, the licensee has an emergency procedure on natural circulation, but cur-rently it is not in conformance to the Emergency Response Guidelines.

The licensee does not yet have NRC-approved Emergency Operating Pro-cedures implemented. The licensee requested on May 5, 1987 an exten-sion in the date for implementation of the Upgraded Emergency Opera-ting Procedures (E0P's); that extension was granted on June 4,1987 by the NRC. The staff agreed that the procedure generation package

.(PGP) which was still under NRC review, should be approved prior to implementation of'the E0P's, which are based on the PGP. The inspec-tor had no further questions.

This Temporary Instruction is closed.

c. Storage Battery Adequacy Audit The NRC:RI Temporary Instruction 87-07 required a review to determine if licensees assure that storage batteries will, in accordance with the current licensing basis, remain properly operable. The NRC will perform an audit to assess the adequacy of control over storage battery operability and ccmpliance with existing NRC requirements.

Attachment 2 of this temporary instruction was previded to the licensee to assist the utility in providing information ef ficiently.

A copy of attachment 2 is appended to this inspection report. (See Attachment 1).

This Temporary Instruction remains open.

d. Fitness for Duty (Drug Testing) Information and Reporting I

The NRC:RI Temporary Instruction 88-01 was issued to collect addi-tional data for the NRC's review of the experience associated with drug testing as part of the licensee's fitness for duty program.

The inspector examined records and data relating to the experience associated with the licensee's fitness for duty program. Information was provided to the NRC:RI as requested in Temporary Instruction

88-01.

l This Temporary Instruction is closed.

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13. Procurement of Zine Conduit Couplings The inspector . receiv'ed _ a copy of a January ,25, 1988 memorandum from W. F. Kane, Director, Division of Reactor Projects, NRC:RI to C. E. Rossi, Director, Division of Operating Events Assessment, NRC:NRR, -which con-cerned fraudulent zinc conduit coupling material used at the Limerick 2 plant. As a result of the NRC:RI concern that other licensees could- be in receipt of r'raudslent coupling material from Franklin Electric Company, the inspecter reviewed the licensee's procurement practices ~ for coupling material. Since the licensee does not procure its conduit or coupling material from Franklin Electric Company, there is reasonable assurance that fraudulent material of the type described is not utilized at the YNPS.

14. Licensee Response to IE Bulletins The licensee's response to the following IE Bulletin (IE8) was reviewed.

This review included: adequacy of the response to IEB requirements, time-liness of the response, completion of identified corrective actions and timeliness of completion.

IE Sulletin No. 85-03: Motor-0perated Valve Common Mode Failures During Plant Transients Due to Improper Switch Settings dated November 15, 1985.

This IE Bulletin requested licensees to-develop and implement a program to ensure that switch settings on certain safety-related motor-operated valves are selected, set and maintained correctly to accommodate the maximum differential pressures expected on these valves during both normal and abnormal events within the design basis.

The licensee responded to the IE Bulletin on May 14, 1986; this response was reviewed by the Division of Operational Events Assessment, NRC:NRR, as documented in a memorandum dated January 26, 1938 issued to the Director, Division of Reactor Projects, NRC:RI.

As requested by action item e. of the bulletin, the licensee identified the selected safety-related valves, the valves' maximum differential pressures and the licensee's program to assure valve operability in its letter dated May 14, 1986, i Review of this response indicates that the licensee's selection of the applicable safety-related valves to be addressed and the valves' maximum differential pressures meets the requirements of the bulletin and that the .

program to assure valve operability requested by action item e. of the t' bulletin is acceptable.

i l

The results of the inspections to verify proper implementation of this program and the review of the final response requird by action iten f.

l of the bulletin will be addressed in additional inspection reports.

l This bulletin remains open.

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15. Participation in NRR/ Licensee Meeting On February 11, 1987, the inspector and the NRC:RI reactor projects sec-tion chief for YNPS attended the monthly meeting between the YAEC licens-ing engineer for YNPS and the NRC:NRR project manager. The meeting in-cluded (1) the licensee's future test plans to meet the requirements of 10 CFR 50, Appendix J, (2) the current status of submitted proposed technical specification changes, (3) the licensee's plans and details associated with the future submission of proposed technical specification changes, (4) the status of various systematic evaluation program items, (5) the licensee's request for an exemption from the implementation of the ATWS rule 10 CFR 50.62, (6) the status of safety issues management system items, (7) the licensee's progress on its program to upgrade emergeacy operating procedures, and (8) other licensing issues of mutual interest to the licensee and NRC.

Subsequent to this meeting, the inspector contacted the NRC:NRR project manager and cognizant NRC:RI management to determine the manner in which the licensee's plans to conduct future 10 CFR 50, Appendix J-required containment integrated leak rate test will be reviewed by the NRC. As a result of these discussions the inspector learned that NRC:RI will review '

the licensee's plans, which are outlined in letter FYR 88-14 dated January 20, 1988. The licensee's schedule for conduct of its next containment integrated leak rate test is an open item (50-29/88-01-03).

16. Unresolved Items An unretolved item is a matter about which more ir. formation is required to ascertain whether it is an acceptable item, a deviation, or a viola-tion. An unresolved item is discussed in Section 8 of this report.

17. Open Items An open item is a matter that requires further review and evaluation by the inspector, including an item pending specific action by the licensee and a previously identified violation, deviation, unresolved item and programmatic weakness. Open items are used to document, track and ensure adequate follow-up by the inspector. Open items are discussed in Sections 4 and 15 of this report.

18. Management Meetings During the inspection period, the following management meetings were con-ducted or attended by the inspector as noted below:

I --

The inspector attended an information meeting held by J. DeVincentis, Vice President and Acting Manager of Operations on February 3, 1988 to discuss items of mutual interest and the NRC inspection program.

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22 The inspector attended an exit meeting held February 11, 1988 with NRC:RI Health Physics specialists at the conclusion of Inspection 50-29/88-04, which consisted of a review of the licensee's Health Physics Program.

The inspector attended an exit meeting held on February 11, 1988 with an NRC:RI physical security inspector at the conclusion of Inspection 50-29/88-03, which consisted of a review of the licensee's physical security program At periodic intervals during the course of the inspection period, meetings were held with senior facility management to discuss the inspection scope and preliminary findings of the resident inspectors.

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r ATTACHMENT-A t i

NRC:RI TEMPORARY INSTRUCTION 87-07 ATTACHMENT 2 STORAGE BATTERY INSPECTION SAMPLE The following identifies the wet cell battery inspection sample. It may be provided to the licensee for more efficient identification of data relevant to l assessing compliance with the current licensing basis.

1. General Battery Information Document the below information for batteries which carry vital loads.

(1) Qualified, or design, seismic life.

(2) Qualified, or design, electrical life.

(3) Age.

(4) Time in service.

(5) Plans for replacement. -

2. Previous Licensee Actions identify actions taken on the following IE Inforr.ation Notices: 83-11, Possible Seismic Vulnerzbility of Old Lead Storage Batteries; 84-83, Various Battery Problems; 85-74, Station Battery Problems; and 86-37, Degradation of Station Batteries.

3. Seismic Lifetime and Qualification For batteries supplying vital loads, identify the following information.

(1) Licensee and/or manufacturer's establishment of seismic lifetime.

This maybe through documentation allowing verification by competent personnel other than the qualifiers and containing design specifica-tions, the qualification method, results, and justifications (ref:

IEEE 535-1986).

(2) Seismic qualification maintenance. Identify how the criteria for assuring that the battery and rack will maintain seismic qualifica-tion are defined, available, and used for periodic inspections and cell replacements. Identify the criteria for determination of l~ seismic end of life based upon the in-service condition of the L battery. .

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4. Electrical Sizing and Qualification For batteries supplying vital loads, identify the following information.

(1) Confirmation that the battery size is sufficient to handle the load profile with a suitable margin.

(2) The means of tracking and control of battery loads such that the batteries and their replacements will have sufficient capacity throughout design life, if worst case electrolyte temperature and other worst case conditions exist when the battery is called upon to perform its design function.

(3) The provisions for consideration of the effect of jumpered out cells upon the ability of a battery to perform under worst case conditions.

5. Battery Ventilation and Protection From Ignition Hazards For batteries carrying vital loads, identify the following.

(1) The provisions for assuring adequacy battery ventilation during normal operation, outages, charging, and discharge.

(2) Adequacy of checks of battery ventilation flow.

(3) Adequacy of contrels over battery ventilation impediments such as enclosing the battery space or its ventilation with plastic sheeting, or any other ventilation obstructions, during outages and other periods.

(4) Adequacy of hydrogen detection equipment and its calibration and use, or of the technical justification for not using such equipment.

(5) Knowledge of the hydrogen hazard on the part of plant management, operating shift management, and personnel who access the battery spaces.

(6) Prohibition of hot work and smoking in battery spaces, including checking the spaces for the residue of such activity.

(7) Assurance that battery cells are secured, with post-to-case and top-to-jar seals tight. Thermometers should not be left in cells after temperatures are measured. Caps on the filler openings should

be properly secured when not required to be off. (Cells should be vented only through the flash arrestors.)

(8) The means of assuring proper elimination of water-carrying pipes (e.g.,

HVAC lines) from battery spaces, especially those which may carry salt water.

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4 (9) The means of positive control over the quality of water added to the batteries to assure that the manufacturer'.s recommendations or an appropriate licensee standard are met or exceeded.

(10) The assurance of elimination of combustibles, and loose equipment and conductors, from battery spaces.

6. Electrolyte Temperature Control For batteries supplying vital loads, identify the adequacy of the following.

(1) Avoidance of localized heat sources such as direct sunlight, radiators, steam pipes, and space heaters.

(2) That the location / arrangement provides for no more than a 5F difference in cell temperature, as confirmed by measurements representative of operating conditions. If this is not the case, then the licensee and manufacturer should have identified the consequent impact on expected battery and individual cell capacity and life, and surveillance procedures should reflect the additional allcwable temperature variation.

7. Charging For batteries carrying vital loads, identify the adequacy of the following.

(1) Provision for a freshening charge after more than 3 months of being on open circuit, unless determined by the manufacturer to be unnecessary to assure rated capacity throughout life.

(2) Accomplishment of equalizing charges at 18 month intervals, and when I

the corrected specified gravity (SG) of an individual cell is more than 10 point (0.010) below the average of all the cells, and when the average corrected SG of all cells drop more than 10 points below the average installation value, and if any cell voltage is below 2.13V.

t (Specific manufacturer's provisions and assessment may allow the non-I performance of some of these recommended charges, or may provide different criteria.)

(3) Control over battery water quality such that specified purify is confirmed before addition, that water added just prior to charging i is added only to bring the electrolyte up to the prescribed minimum i

(to prevent overflow during charging), and that water added af ter and between charges not bring the level above the prescribed maximum (unless manufacturer's instructions provide for other water addition measures).

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.e, (4) That routine ficat and final end of charge SGs not be taken before 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of float operation after completion of the charge and 'the last water addition, unless the manufacturer's instructions provided otherwise. (The need is for measurement of representative cell levels and average them.)

(5) Establishment and maintenance of float voltage on accordance with the manufacturer's' instructions.

(6) Assurance that single-cell charger use does not violate Class 1E independence from non-class 1E equipment.

8. Performance Tests and Replacement Criteria For batteries carrying vital loads, identify the following.

(1) Initial acceptance testing which demonstrates the ability of the batteries to meet the manufacturer's rating.

(2) Service testing which demonstrates the ability to carry the load profile with an appropriate margin for worst case conditions, including end of life loss of capacity under the worst case electrolyte temperature.

(3) Accomplishment of a performance test (capacity test discharge) within the first two years of service and at 5 year intervals until signs of degradation are evident or 85'e of the qualified service life is reached.

(4) Annual performance testing of batteries which show signs of degradation or which have reached 85'; of the qualified service life is reached.

(5) End of electrical life criteria which consider the rapid end of life drop-off in capacity, worst case state of charge during float service, worst case electrolyte temperature, current DC loads, and the time needed to replace the battery while it can still handle worst case conditions, j 9. Other Safety-Significant Wet Cell Batteries i

For safety-significant wet cell batteries not used for vital loads, show

, how the maintenance program periodically determines the ability to l perform the design function and provides for timely replacement of l batteries and for maintaining associated equipment (e.g., chargers).

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