IR 05000309/1996016

Insp Rept 50-309/96-16 on 961208-970128.No Violations Noted. Major Areas Inspected:Operations,Maint,Engineering & Plant Support

ML20134M521
Person / Time
Site:	Maine Yankee
Issue date:	02/11/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20134M502	List: IR 05000309/1996016 ML20134M499
References
50-309-96-16, NUDOCS 9702200277
Download: ML20134M521 (36)

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Enclosure 2 U. S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket No: 50-309 License No: DPR-36 Report No: 50-309/96-16 Licensee: Maine Yankee Atomic Power Company (MYAPC)

Facility: Maine Yankee Atomic Power Station Location: Bailey Point Wiscasset, Maine Dates: December 8,1996, through January 28,1997 Inspectors: Harold Eichenholz, Project Engineer Division of Reactor Projects Jimi Yerokun, Senior Resident inspector Division of Reactor Projects Kenneth Kolaczyk, Reactor Engineer Division of Reactor Safety Approved by: Richard Conte, Chief, Projects Branch No. 5 Division of Reactor Projects 9702200277 970211 PDR ADOCK 05000309 G PDR

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EXECUTIVE SUMMARY - 1

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Maine Yankee Atomic Power Company l

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NRC Inspection Report 50-309/96-16 )

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A special inspection was conducted to review the status of issues identified by the NRC I ISA team as documented in their report dated October 7,1996. The inspection was  !

i focused on ensuring that issues are properly resolved by the NRC. Some of those issues j were addressed in previous NRC inspection reports (50-309/96-09,96-10, or 96-11). The

details section of this report addresses all the other issues not previously discussed in the above enumerated reports. Also for clarity purposes, inspection findings involving l apparent violations identified in the four (4) subject reports are grouped into specific areas

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of concern. Therefore, based on the results of these inspections, the following issues were identified as deficiencies and considered to be apparent violations of NRC

requirements

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}- In the area of safety related equipment inoperability, six (6) apparent violations were

! identified due to design related and configuration control problems. These were: an

, inoperable high pressure safety injection (HPSI) pump due to a cut wire (TS 3.9)

l [50-309/96-11, [Section M4.11; an inoperable control room ventilation and filter system

, due to the inability to maintain a positive pressure within the control room (TS 3.25.B.2), .

i [Section E3.2b(11)}; thirty (30) examples of electrical equipment important to safety that

were not qualified for service in a submerged environment as a result of post accident -1

conditions (10 CFR 50.49, Section b.3 and Section e.6) [50-309/ 96-10, Section E2.2]; a ,

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design deficiency involving non safety-related air operated dampers in the containment

, spray building's ventilation system that had the potential to result in inoperability of i emergency core cooling equipment (10 CFR 50, Appendix B, Criterion Ill) [50-309/96-09,

! Section E2.1]; and inadequate overpressure protection for sections of the primary component cooling (PCC) water system that had potential, with the application of single

[ failure conditions applied to the secondary component cooling (SCC) water system, to  !

l render both trains of emergency core cooling systems (ECCS) and residual heat removal l l heat exchangers inoperable (TS 3.6), and/or both emergency diesel generators (EDGs)

inoperable (TS 3.12), [Section E2.1.c].

In the area of testing inadequacies, five (5) apparent violations were identified due to TS i coverage or procedural problems. These were: nine (9) instances of failure to test j instrumentation and control circuits of the safety injection actuation system, the main j steam isolation valve control, the feedwater trip system, emergency feed water (EFW) l l initiation, and refueling water tank level recirculation actuation system initiation (TS 4.1) I

! [50-309/96-11, Section E3.2]; three instances (3) of failure to perform emergency power j system periodic testing (TS 4.5) [50-309/96-11, Section E3.2]; a failure to perform j periodic testing of the feedwater trip system (TS 4.6) [50-309/96-11, Section E3.2); six l (6) instances of failure of the test program to assure that safety related equipment will j perform satisfactorily contrary to 10 CFR 50, Appendix B, Criterion IX involving:

The EDG room fan thermostats [50-309/96-09, Section M3.2]; the PCC and SCC water j i system's flow control valves [50-309/96-09, Section E2.2]; EDG load sequencers, the j motor driven fire pump, permissive relays for the trip block timers used for the low pressure safety injection (LPSI) pumps, for instance [Section M1.2b(2)(l)]; and control

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board annunciator fault alarm circuits in the ECCS system, [Section M1.2] (10 CFR 50, Appendix B, Criterion XI); and (15) instances of inadequate inservice testing for the pump discharge check valves for the EFW, HPSI, LPSI, PCC, SCC, and service water pumps, '

Section M1.2b(2)(lll)) (TS 4.7/10 CFR 50.55a). These deficiencies represent a significant weakness in the area of testing, which resulted in a number of apparent violations. Also, the weakness in this area was a contributor to two (2) of the safety related equipment inoperabilities discussed above (i.e., the HPSI pump inoperability due to a cut wire and the inability to maintain a positive pressure in the control room).

In the area of safety review, two apparent violations reflecting inadequate safety evaluations were identified. The first apparent violation involves three instances of failure to perform adequate safety evaluations in accordance with 10 CFR 50.59(b)(1) for changes ,

made to procedures and the facility as described in the Updated Final Safety Analysis )

Report (UFSAR). These included: instances involving the inability to support plant I operations up to the service water temperature values stated in the UFSAR [Section E3.2(b)(2); the development of a safety evaluation that erroneously concluded that the

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inability to maintain positive pressure in the control room by the control room breathing air l supply system while the plant was being operated was not an Unreviewed Safety Question ,

[Section E 3.2b(11)]; and the failure to perform a proper safety evaluation prior to !

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developing a procedure that permitted the cross-connecting of DC buses, an operation prohibited by the UFSAR, [Section E 3.2b(8)). The second apparent violation involves numerous changes that have been made to the FSAR without providing annual updates to the UFSAR (10 CFR 50.71(e)(4), [Section E3.2b(S)}. In addition to the failure to perform

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adequate safety evaluations, as described above, a Maine Yankee initiative to upgrade the UFSAR has identified numerous changes that may need a 10 CFR 50.59(b)(1) safety l i evaluation to support changes made to procedures or the plant that are different than '

described in the UFSAR [Section E3.2b(9). Accordingly, at the enforcement conference,

Maine Yankee should be prepared to address the existence of additional but unknown deficiencies in this area.

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In the area of procedure inadequacies and non-adherence, two (2) apparent violations were I identified. The first apparent violation involved three (3) instances of failure to establish (have adequate) rrocedures as required by TS 5.8.2. In the first instance, plant procedures for certain emergency events would have precluded the maintenance of minimum control i room staffing as required by the TS, [Section 04.11. The second and third instances I involved the f dfure to establish maintenance procedures for the conduct of a vendor )

recommended examination on a component of the "A" EFW pump and the requirement to install and control fastener lockwire on safety related components, [Section M1.1]. The second apparent violation involved three (3) instances of failure to implement procedures.

These instances involved the failure to properly maintain control room logs [Section 04.2), I to properly conduct an operability review [Section 04.3], and to improperly remove a

] seismic support on a safety related service water pump [Section M1.1].

In the area of corrective actions not identified, untimely, and/or inadequate, an apparent violation (with multiple instances) 10 CFR 50, Appendix B, Criterion XVI was identified

[Section E8.11. Regarding the failure to identify problems that needed to be in the corrective action system, three (3) instances were identified: the existence of a design deficiency (i.e., the containment spray building ventilation design that relied on a non safety-related source of instrument air); the failure to identify components in containment l iii

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that were below submergence level in the Environmental Qualification Program; and the failure to recognize the lack of complete testing of the ECCS actuation logic. Instances of untimely corrective actions involved the following two (2) examples: the design deficiency associated with the dampers for the containment spray building ventilation system, and an identified issue involving turbine hall flooding that puts the plant outside its design basis.

Two instances of ineffective corrective actions were identified and involved the auxiliary -

feedwater control system and the manner in which the negative control room pressure

~ identified during a surveillance test was addressed. The first instance also reflected repetitive problems.

In addition, thirty one (31) items were identified as unresolved because further reviews are needed by the NRC to determine their proper disposition. Those items arn listed as attached at the end of this report.

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TABLE OF CONTENTS TAB LE O F CO NT ENT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v l . O p e r a tio n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

, ......................................................... 1 04 Operator Knowledge and Performance . . . . . . . . . . . . . . . . . . . . . . . . . 1 04.1 Operations Assessment, Control Room Staffing . . . . . . . . . . . . . . 1 04.2 Control Room Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 04.3 Operability Determinations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3

11. M a i nt e n a n c e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 M1 Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 M1.1 Quality of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 M1.2 Testing (URI 50-309/96-16-04 through 50-309/96-16-06) . . . . . . 5 M2 Maintenance and Material Condition of Facilities and Equipment . . . . . . . 9  !

M2.1 Plant Walkdown Observations (URI 50-309/96-16-07) . . . . . . . . 9

111. Engineering .................................................. 10 E2.1 Followup to July 19,1996 Plant Shutdown ............... 10

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E3 Engineering Procedures and Documentation . . . . . . . . . . . . . . . . . . . . 13 E3.1 Design and Licensing Bases Discrepancies (URI 50-309/96-16-

, 08, through 50-309/9 6-16-15 . . . . . . . . . . . . . . . . . . . . . . . . . 13 E3.2 FSAR Discrepancies (URI 50-309/96-16-16 through 50-309/96-i 16-24).......................................... 15 E3.3 Design / Engineering Related issue (URI 50-309/96-16-25

thro ug h 2 7) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 E8 Miscellaneous Engineering issues ........................... 22

E8.1 Corrective Actions (URI 50-309/96-16-28 through 30) . . . . . . . . 22 I V. Pl a n t S u p p o rt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 i R1 Radiological Protection and Chemistry (RP&C) Controls . . . . . . . . . . . . 24 R1.1 Hadiation Protection Program (URI 50-309/96-16-31) ........ 24 i V. Ma nagem ent Meeting s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 j X1 Exit Meeting Summ ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 i I

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PARTIAL LIST OF PERSONS CONTACTED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25  !

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INSPECTION PROCEDURES USED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 ITEMS OPENED, CLOSED, AND DISCUSSED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 LIST OF ACRONYMS USED ......................................... 30 l v

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Report Details

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Purpose

The purpose of this inspection was to ensure that selected safety concerns identified by ' i

. the NRC Independent Safety Assessment (ISA) team had been resolved and to assure

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proper regulatory disposition on the issues, and to review and vsrify the actions taken by your staff in response to the issues. The feam's issues were documented in their report

. which was transmitted to you on October 7,1996. This inspection consisted of reviews

of the ISA team's report, selected examinations of design documentation, procedures and

. representative records, personnel interviews, and review of the actions taken by your staff I to address selected issues.

Background i

in 1996, the ISA team performed an evaluation of Maine Yankee's safety performance.

{ The team independently assessed Maine yankee's conformance to its design and licensing basis; t,perational safety performance; effectiveness of licensee self-assessments, corrective actions, and improvement plans; and determined the root cause(s) of safety

]' significant findings. The results of the team's inspections were documented in a report transmitted to Maine Yankee on October 7,1996. )

l The team concluded that while overall performance at Maine Yankee was adequate for i i operation, a number of significant weaknesses and deficiencies existed. The NRC Regional

! and Headquarters staff reviewed the team's report and also Maine Yankee's actions to

, address the identified discrepancies. During the ISA team's inspection, certain items with

! immediate safety significance were identified as restart items. Those items were inspected

}~ for adequacy and the results docamented in NRC Inspection Report 50-309/96-08. Other items requiring immediate corrective actions wero reviewed for adequacy and the results documented in NRC Inspectbn Report 50-309/96-09. The issues associated with the Environmental Qualification (EO) of equipment in the containment were inspected and documented in NRC Inspection Report 50-309/96-10. The issues associated with the Engineered Safeguards Features (ESF) logic testing were inspected and documented in NRC Inspection Report 50-309/96-11. All other issues identified in the ISA team's report were subsequently reviewed and are being dispositioned in this report.

i 1. Operations 04 Operator Knowledge and Performance i 04.1 Operations Assessment. Control Room Staffina i

a. Inspection Scope (71707,92901)

i In Section 3.1.1.2 of the ISA team report, it was noted that Maine Yankee had I established conflicting procedural requirements that would be impossible to comply l with by Control Room (CR) operators in the event of a fire coincident with a medical

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emergency. During this postulated event, the minimum CR staffing requirements i

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required in Technical Specification (TS), Section 5.2.2/ Table 5.2-1, would not be satisfied if the senior reactor operators responded as described in MY's i I administrative procedures. l b. Observations. Findinos and Conclusions  !

i Technical Specifications 5.8.2 requires that aopropriate procedures (adequate) are l established and properly implemented. Plant procedure 1-200-10, Conduct of )

l- Operations, Section 4.6, described the Fire brigade requirements which called for !

l the Shift Operating Supervisor (one of the two required senior reactor operators) to l l be the leader of the Fire Brigade. Also, procedure 1-26-4, Responsibilities and i Authorities of Operating Personnel, described one of the responsibilities of the Plant i Shift Superintendent (the second of two required senior reactor operators) as l responding to a medical emergency. As of October 7,1996, MY would have failed to meet these procedural requirements assuming that Technical Specifications the minimum CR staffing requirement were satisfied in the event of a fire coincident with a medical emergency and therefore this item is considered a violation (Apparent )

Violation of Technical Specification 5.8.2). Additionally, the inspector noted that j the occurrence of the subject event during backshift operations, could preclude the l Plant Shift Superintendent from fulfilling the Interim Emergency Director's duties specified in the Maine Yankee Emergency Plan. Corrective measures were implemented by Maine Yankee to address the staffing concerns. The measures included turning over the responsibility for responding to a medical emergency to the security department.

04.2 Control Room Observations a. Insoections Scooe (71707. 92901)

The ISA team noted in Section 3.1.2.1 of the Report that control room logs lacked detail and did not meet the guidance provided in MY's administrative procedures.

The team determined that operators did not log the starting time and stopping time of an EDG, equipment problems experienced during the plant shutdown, and the ,

initiation of a manual reactor trip when problems were experienced with the control !

element assemblies drive system. l l

b. Observations. Findinos and Conclusions Administrative procedure No.1-200-3, Rev. 5, Operations shift Records and Logs, i specifies in Section 3.2.7, Control Room Log Book, that the following types of entries shall be recorded in the control room log book: major equipment status change and special plant conditions or observations. Technical Specification 5.8.2 requires that plant procedures be implemented. The failure of control room l operators to adequately enter plant operational conditions into the control room log

! as required by procedure 1-200-3 is considered a violation (Apparent Violation of l Technical Specifications 5.8.2).

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l a. Inspection Scope (71707. 929011 ' I

According to Section 3.1.2.5 of the ISA team report, an incorrect interpretation of '

TS requirements involving logic testing as required by TS Table 4.1-2, Minimum l Frequencies for Checks, Calibrations and Testing of Engineered Safeguards Systems i instrumentation and Controls occurred. l

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b. Observations. Findinas and Conclusions  !

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in this case, a correct operability determination for inoperable HPSI pump and containment spray swing pump was initially made by the operating shift, in accordance with procedure 1-200-10, Conduct of Operations; but that decision was '

! subsequently reversed. Following further discussions between MY and the NRC, a

clarifying memorandum was issued by the Operations Manager to operating l personnel stating that if a safety related logic circuit testing deficiency was i i identified, the associated components would be considered inoperable due to a I failed surveillance and the appropriate TS would be entered. In MY's December 10, I

1996 response letter to the ISA Report, Enclosure 7 - Clarification of Statements in '

the ISA Report, stated that MY disagreed with the NRC position.  ;

i 3 The failure of MY to perform an appropriate operability determination as required by ,

! section 4.13.5 of procedure 1-200-10, Conduct of Operations, is considered an ,

j example of a violation of failure to follow procedure. (Apparent Violation of  :

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Technical Specification 5.8.2).

i 04.4 Technical Soecification Interoretations (URI 50-309/96-16-01)

l_ a. Insoection Scope (71707,92901)

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I As a result of reviewing MY TS interpretations, two concerns were identified by the

ISA team in Section 3.1.2.6 of the ISA Report. First, the team identified that MY j . was incorrectly applying the TS 4.0.A provision that allows for an extension of +25 l percent for surveillance time intervals to audit intervals. The second concern was

that an interpretation in use at MY since 1985 allowed for the use of a daily containment grab sample as one of the two TS 3.14 required primary coolant leak i detection systems.

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b. Observations. Findina.and Conclusions

! For the first concern, during this inspection, MY was requested to provide' an

assessment of the impact that their implementation of this incorrect interpretation i

had on TS 5.5.B.9 compliance. A MY representative reported that in 1994,

contrary to TS 5.5.B.9.h, the requirement to perform an annual fire protection and j loss prevention inspection and audit was exceeded by a 2-day period. Also, the TS .

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5.5.B.9.c requirement to audit once per 6-months the results of actions taken to

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of operation that affect nuclear safety was exceeded by 9-days in 1992. In the

! summer of 1996, MY canceled the interpretation.

Based upon the above information, the failure of MY to meet the periodic audit -

frequencies established by TS 5.5.B.9 is considered a violation. However, this failure constitutes a violation of minor significance and is being treated as a non-cited violation, consistent with Section IV of the enforcement policy.

In the second concern, MY identified that the leak detection system was sensitive to radioactivity. However, the ISA team concluded that this condition was a change to a TS because it did not represent a continuously operating system. At the conclusion of the assessment, MY disagreed with the team, and it was concluded that further NRC review would be required to determine the appropriateness of this MY TS interpretation. Therefore, this item remains open pending completion of NRC ;

review. (URI 50-309-96-16-01). j

04.5 Post Trio Reviews (URI 50-309/96-16-02)

a. inspection Scooe (92901)

in Section 3.1.2.7 of the ISA Report, the team noted that MY performance in the area of performing post trip reviews (PTR) was weak, in that they lacked rigor and completeness.

b. Observations. Findinas and Conclusions One PTR indicated the unavailability of the plant process computer. The weak performance in the manner in which PTRs are developed by MY and the issue of unavailability of the plant process computer is considered unresolved pending further NRC review (URI 50-309/96-16-02).

04.6 Emeraencv Ooeration Procedures (URI 50-309/96-16-03)

a. Insoection Scope (92901)

In sections 3.1.3.1 and 2.4.2 of the ISA Report, a concern involving the inability to isolate an affected steam generator following a steam generator tube rupture (SGTR)

event within the time stated in the UFSAR was identified.

b. Observations. Findinas and Conclusions it is noted that prior to compensatory measures being put in place by MY to add;'ess this issue (i.e., providing additional operator training and establishing an administrative limit on reactor coolant system activity), the potential existed to overfill the steam generators thereby releasing radioactive liquids to the

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environment, possibly exceeding the limits of 10 CFR Part 100. The NRC is l currently reviewing MY's actions to perform 50.59 reviews in this area, including

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the development of the safety analysis associated with establishing the more restrictive limits for reactor coolant activity. This item remains open pending completion of NRC's reviews (URI 50 309/96-16-03).

II. Maintenance M1 Conduct of Maintenance M1.1 Quality of Maintenance a. Insoection Scoos (62707,92902)

In Section 3.2.3 of the ISA Report, three examples of poor work practices were identified by the team.

b. Observations. Findinas and Conclusions The examples of poor work practices involved:

(1) Failure to perform a vendor recommended magnetic particle inspection of used emergency feedwater (EFW) pump diffusers prior to their reassembly as part of a 1995 overhaul of the P-25A EFW pump.

(2) Improper removal of a seismically qualified pipe support on a seal water line for SW pump P-29C by maintenance personnel on August 13,1996 without tagging out the pump er declaring the pump inoperable. This was contrary to the applicable work order and Section 6.6 of 0-16-3.

(3) Lack of procedural detail for installation and control of fastener lockwire on six (6) safety related components.

These instances of poor work practices were contrary to the requirements of procedure 0-16-3, Work Order Process, which prescribed the process and controls for the coordination, implementation, and documentation of maintenance activities.

As a result, these instances were examples of failure to meet the requirements of Technical Specification 5.8.2, which requires that plant procedures be developed (ltems (1) and (3) above) and be implemented (Item (2) above) (Apparent Violation of Technical Specification 5.8.2).

M1.2 Testina (URI 50-309/96-16-04 throuah 50-309/96-16-06)

a. Insoection Scooe (61726,92902)

in section 3.2.4 of their report, a number of issues and deficiencies identified by the ISA team involved activities controlled by MY's Testing Program and the adequacy of related test procedures.

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b. Observations. Findinas and Conclusions Follow up of issues in'this area in this inspection identified that certain testing activities were either not performed at all, lacked rigor, or were not consctly performed. These were:

(1) HPSI Flow Testina and Throttle Valve Settinas in Section 2.2.1.2 of the ISA Report, the team identified concerns with lack of rigor (i.e., the adenuacy of documentation and instrumentation) during HPSI pump testing at high flow conditions in 1993. To address NRC concerns in this area, MY plans a future test of the HPSI pumps in a technically rigorous manner to fully demonstrate the available margin. This testing will also afford MY the opportunity to reset the position of the system's throttle valves, and therefore address the issue discussed in ISA Report Section 2.2.1.3. This issue involves ensuring that the position of the system's throttle valves will use a more precise tolerance to ensure that required flow is met and pump runout conditions are not exceeded. The NRC Region I will provide further review of this testing activity and the manner in which the throttle

. valves are reset, therefore this matter will remain an unresolved item (URI 50-309/

96-18-04).

(2) Testina Weaknesses in Demonstratina the Functionality of Safetv Eauioment As documented in ISA Report Section 3.2.4, the team found inadequacies in the scope of testing programs, and weaknesses in the rigor in which testing was performed, and in the evaluation of testing results to demonstrate functionality of safety equipment. Examples of the individual inadequacies and weaknesses are:

(l) Poor Emeroency Diesel Generator (EDG) Testina Procedures.

.The ISA team identified that safety-related electrical time-delay relays associated with the EDG's load sequencer were not verified for proper operation and were not in MY's calibration program. This was evident from the fact that no tolerance band and acceptance criteria had been established and the relays had not been calibrated since installation. What testing is performed on load sequence timer relays (e.g.,

62-5,62-P25A, and 62-P25C) is controlled by procedures 3.1.14A/B, "A/B" Train EDG/ECCS Cold Shutdown Test. A further deficiency was identified by the ISA team, in that, Step 5.2.6 of procedures 3.1.14A/B was incorrect because the sign off for the 20 second start block for P-25C/A was listed in the procedure as 30 seconds, was performed many times and the 10 second error in the timing sequence was never questioned.

Subsequently, MY identified two additional testing inadequacies involving relays that were neither tested nor included in their calibration program. These were, the motor-driven fire pump start permissive relay 62-P4 (Note, MY's 10 CFR 50, Appendix B QA Program is applied to fire protection features) and the safety-related permissive relay 62-RAS to remove the low pressure safety injection (LPSI) pump

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trip 10 seconds after the recirculation actuation signal to allow manual restart of the pump. For the latter function, procedure 3.1.15.2 did not verify the trip block function of the 62-RAS timers.

These four instances of failure to perform proper testing and/or calibration of safety- ;

related or fire protection equipment was an apparent violation of 10 CFR 50, Appendix 0, Criterion XI, Test Control, which requires that a test program to assure that all testing required to demonstrate that structures, systems, and components :

will perform satisfactorily in service is identified and performed in accordance with written test procedures (Apparent Violation of 10 CFR 50, Appendix B, Criterion XI).

(11) Control Room Ventilation Test Results Not Properly Evaluated The follow-up inspection in this area identified inadequacies in the area of 10 CFR !

50.59, and as such, the inspection findings are discussed in Section E3.2.

(Ill) Inaoorooriate inservice Test Procedures For Check Valve Testina  !

The ISA team questioned the use of a 0-3000 psig gauge for a 15 psig measurement during the testing of a check valve in the emergency feedwater ;

system. In addressing this discrepancy, MY Technical Evaluation No.142-96, dated August 27,1996, identified 14 other deficiencies in test methodology for the folowing components / procedures:

Chg/HPSI Pp Disch Cks CH-10,19,26 Procedure 3.1.2.4 Otriy test i Elec. EFW Pp Disch Cks EFW-15,314 Procedure 3-1-22 Cold S/D test LPSI Pp Disch Cks LPSI-50,51 Procedure 3.1.20.2 Otriy test i l

PCC Pp Disch Cks PCC-6,13 Procedure 3.1.2.8 Otriy test  !

l SCC Pp Disch Cks SCC-7,14 Procedure 3.1.2.7 Otrly test Svc Wtr Pp Disch Cks SW-1,4,7,10 Procedure 3.1.2.9 These 15 instances of inadequacies were contrary to the requirements of Technical Specification 4.7, to establish an Inservice Testing Program per 10 CFR 50.55a (Apparent Violation of Technical Specification 4.7).

The inspector reviewed the Licensee Event Report (LER 96-28), dated August 28, 1996 written to address these inadequacies. Details were consistent with the Technical Evaluation (TE 142-96).

(IV) The inadeauste testina associated with emeroency actuation circuits were addressed in NRC Inspection Report 50-3t* ~ 3-11 and identified as aooarent violations of technical specification 3.9-2, 4.1. 4.5 and 4.6

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(V) Control Board Annunciator Fault Alarm Circuits Not Periodically Tested l

The ISA team identified four annunciator fault alarms in the safety-related ECCS l system that were neither periodically tested nor had an established test procedure presenbing the necessary instructions to perform testing of these circuits.

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l l The following alarm response procedures cover the alarms in question, and I corresponding testing procedures that are identified are those procedures that failed l

to provide the requisite testing: I

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AOP 2-37.RH Page 35 of 52, SIAS 86 Device Trip Path Fault / Proc. No.

3-6.2.1.5.4, Safeguard Channel Calibration Safety injection Actuation Signal.

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AOP 2-37.RH Page 41 of 52, RAS 86 Device Trip Path Fault / Proc. No. J 3-6.2.1.41, Indication for Safeguard Channel Cerbration for RWST.

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AOP 2-37.RH Page 48 of 52, CIS 86 Device Trip Path Fault / Proc. No. 1 3-6.2.1.5.2, Safeguard Channel Calibration Containment isolation Actuation ;

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3-6.2.1.5.3, Safeguard Channel Calibration Containment Spray Actuation Signal.

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These four instances of failure to perform testing and develop a test procedure for safety-related equipment, as enumerated above, was contrary to the requirements of 10 CFR 50 Appendix B, Criterion XI, Test Control, which requires that a test program to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures (Apparent Violation of 10 CFR 50, Appendix B, Criterion XI). j (Vt) Standbv Power Meters Reauired Bv R.G.1.97 Not Calibrated And Periodically Tested Certain meters that woi be used in a post accident situation were not covered by 10 CFR 50 App. B or IW . QA Plan. This item remains open pending further inspection and/or determining basis that there is a requirement to test / calibrate.

(URI 50-309/96-16-05).

(Vil) Lack of calibration of the Emeroency Diesel Generator Room Fan thermostats was addressed an NRC Insoection Report 50-309/96-09 and Identified as aooarent violation of 10 CFR 50. Apoendix B Criterion XI (Vill) The inadeauate testino of the recirculation actuation sianal circuitry was addresse' vith item IV ebove i l

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(IX) Importance of Air-Ocerated Valve Testina Recentiv Recoanized l

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l 9 l Equipment problems were noted in this area. The last NRC IST inspection in this l area was documented in IR 93-13 (6/28-7/2/93). The issues raised by ISA team ,

warrants a Specialist review therefore, the adequacy of MY IST Program and/or

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10 CFR 50 Appendix B Criterion XI program remains an open item pending l completion of NRC's inspection. (URI 50-309/96-16-06).  ;

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M2 Maintenance and Material Condition of Facilities and Equipment M2.1 f'!gnt Walkdown Observations (URI 50-309/96-16-07)

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i a. Inspection Scope (62707. 92902)  ;

in section 3.2.1.5 of the ISA report, the team discussed some noteworthy l observations that were made during plant walkdowns. The inspector reviewed  :

these observations and plant conditions to ascertain that allidentified adverse I conditions had been corrected. ,

b. Observations Findinas and Conclusions .

t The first issue involves primary component cooling (PCC) water piping corrosion, f The team noted that the physical appearance of the pipe inside the containment appeared severely corroded raising the question of the adequacy of the pipe integrity. This issue was addressed in inspection Report 96-08 and as such, this .

matter warrants no further discussion in this report.

A second area of concern was identified in the service water (SW) pump bay of the circulating water pump building. Extensive material condition problems were  ;

identified, such as water on the floor, corroded fasteners, corroded supports, pump j base plate corrosion, missing u-bolt hanger parts, and cracked grout pads.  !

Additional problems were subsequently identified by MY once their focus was I directed to the area by the ISA team. I I

l A third area of concern identified by the ISA team involved ineffective inspection of the containment for cleanliness after the outage. This item deals with failure of MY )

to eliminate inappropriaM % reign material or items left over from maintenance activities. Examples inclu"ed: items wrapped in unqualified plastic snd discarded tape. Following the team identifying this as an area of concern, MY Identified uncontrolled tools in the containment. This item remains open pending completion of NRC's review of the as-found conditions in the SW pump bay, and the adequacy of MY's inspection of the containment at the completion of an outage.

(UR150-309/96-16-07).

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111. Engineering E2 Engineering Support of Facilities and Equipment E2.1 Followuo to July 19.1996 Plant Shutdown a. Insoection Scoce (92903)

The inspector reviewed the circumstances associated with the Maine Yankee discovery that sections of the Primary Component Cooling (PCC) system piping did not have adequate overpressure protection. To perform the review, the inspector interviewed personnel, reviewed Licensee Event Report (LER)96-022, and PCC system design documents.

b, Observations and Findinas Backoround

On July 19,1996, the Maine Yankee engineering staff identified a design deficiency in the PCC water system. Specifically, during a postulated design basis accident which results in a containment spray actuation signal (CSAS), the expected isolation of PCC flow to the nonsafety-related containment air recirculation (CAR) fan coolers could cause a rupture of the PCC piping inside containment because of heat up of the fluid trapped inside the piping. Since a rupture of the PCC piping could render containment integrity and portions of emergency core cooling systems (ECCS)

inoperable, Maine Yankee operations personnel entered the remedial actions of Technical Specification (TS) 3.6, " Emergency Core Cooling System," TS 3.11 (Containment integrity) and commenced a plant shutdown.

Maine Yankee reported this design deficiency to the NRC on July 19,1996, per the requirements of 10 CFR 50.72 (b)(1)(ii)(B) as a condition outside the plant design basis. To correct this condition, Maine Yankee installed relief valves on the sections of PCC system piping that were susceptible to overpressurization. Details concerning the July 19,1996, plant shutdown and relief valve installation are documented in NRC inspection report 50-307/96-08.

As a result of Maine Yankee's discovery and NRC findings at other utilities that indicated this design deficiency may exist at other plants, the NRC issued information Notice 96-49 on August 20 and Generic Letter 96-06, " Assurance of Equipment Operability and Containment Integrity During Design Accident Conditions" on September 30,1996. Generic Letter 96-06, requested utilities to, in part, examine piping systems that penetrate containment and determine if they are susceptible to thermal expansion of fluid wherein overpressurization of piping could occur. If a utility identified susceptible piping systems during the review, the Generic Letter requested the utility to assess the operability of those systems.

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I Maine Yn@ge Discoverv

- In preparation for the independent safety assessment team (lSAT) inspection, Maine Yankee contracted the Yankee Atomic Electric Company (YAEC) to perform an independent re' view of the PCC system. During the design review in July 1996, a contractor who had been hired by YAEC to assist in the review affort identified the PCC design deficiency.

The contractor noted the PCC piping was susceptible to overpressurization while !

reviewing a June 25,1996, Westinghouse Nuclear Safety Advisory Letter (NSAL) ,

that described design issues associated with CAR coolers at Westinghouse plants. '

The topic discussed in the NSAL did not directly address overpressurization of CAR 1 system piping due to system isolation, the deficiency that was ultimately discovered i at Maine Yankee. The contractor stated the process that he used to review the j NSAL, and examinations of pipe schematics etc. led him to the discovery that sections of Maine Yankee's PCC system piping inside containment did not have relief protection and could be overpressurized during an accident, l

Maine Yankee and YAEC personnel reviewed PCC design basis documents and I discussed the issue with Stone and Webster Engineering Corporation (SWEC), the l constructor of the plant, to determine why design engineers never installed relief J valves on the PCC system piping during initial plant construction or during subsequent system modifications. In a July 26,1996, meeting, Maine Yankee, ]

j YAEC and SWEC personnel reasoned that original plcnt designers had excluded l overpressure protection based upon a belief that during an event, boundary valve i leakage and pipe ductility would limit pressure buildup. Further, the "one-time" loading created by the post-accident heatup was thought by the original designers to represent secondary stresses which did not create credible fatigue or failure concerns. The design basis of the Maine Yankee PCC piping addresses so-called

" occasional loads" in the context of piping standard B31.1 (later termed faulted conditions in more modern piping design codes), but only considered seismic loads in this analysis and erroneously omitted the pressurization in question, in retrospect, as described in GL 96-06, this interpretation was not correct.

NRC Followuo Notwithstanding the document search and interviews performed by Maine Yankee and YAEC personnel, the inspector conducted an independent review of PCC design information to determine why relief valves were not installed on the PCC piping.

The inspector's search involved reviewing portions of the Maine Yankee Final Safety Analysis Report and the Individual Plant Examination Report, PCC design documents, test procedures, and inter office and intra office memoranda.

Additionally, the inspector interviewed Maine Yankee and YAEC engineering and l maintenance personnel. i i

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i None of the documents discussed a need to protect sections of the PCC system j piping from the effects of overpressurization. Further, none of the personnel l

interviewed could recall an instance where an individual had questioned the ,

adequacy of the overpressure protection for the PCC system. The PCC piping at :

Main Yankee was constructed according to the 1967 edition of American National !

Standards institute (ANSI) Standard B31.1., " Power Piping." Although Section !

101.4.2, " Fluid Expansion Effects," requires design engineers to consider where expansion of a fluid may increase pressure, the post accident heatup and '

pressurization load was apparently overlooked.  ;

As part of the follow-up to this event, the inspector reviewed LER 96-022, dated !

August 19,1996, which summarized Maine Yankee's corrective actions following !

the July 19,1996, plant shutdown. Maine Yankee identified the absence of thermal j relief valves on the PCC system as a condition that placed the plant outside of the 1 technical specifications (TS). The inspector verified LER 96-022 was timely, the ;

information appeared to be accurate, and the LER had adequately described the consequences that could be associated with a lack of relief valve protection on the PCC system. l Insoector Assessment i

The PCC system at Maine Yankee supplies cooling water to one train of ECCS i components and one emergency diesel generator (EDG). The other EDG and ECCS  !

train is cooled by the secondary component cooling (SCC) water system, in LER 96-022, Maine Yankee stated that the overpressurization and rupture of the )

PCC piping could result in the loss of several TS-required components. The ,

significance of the loss depended on which component failures and accident ]

i_ sequences were postulated, and what operator actions could be taken to mitigate :

the event. For example, in one scenario, Main Yankee concluded rupture of the l PCC piping during an event, coupled with a failure of the SCC system, could result i in a loss of cooling water to components in both ECCS trains and both Emergency l Diesel Generators (EDGs). Such a scenario, absent operator actions, could  ;

ultimately lead to a loss of post-accident heat removal. In another scenario, Maine !

Yankee indicated rupture of the PCC piping (the second isolation " barrier"), coupled 1 with a failure of the PCC inlet or outlet isolation valves PCC-222 or PCC-A 238, could result in a loss of containment integrity. )l The inspector noted the situations outlined in LER 96-022, were worst-case

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scenarios ano it is possible the PCC system piping would not overpressurize and 4 rupture during an event. Specifically, the PCC penetration isolation valves for tne ;

CAR coolers, check valve PCC-222 and air operated globe valve PCC-A-238, arenot subject to periodic seat leak tests per 10 CFR 50 Appendix J. Therefore, both ;

valves maf have seat leakage. Seat leakage past either valve during an event would !

' limit PCC system pressure buildup. Conversely, if there was not appreciable j l

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seat leakage past either valve, es the trapped water in the PCC system expanded, the isolated portion of PCC piping would also expand in response to the pressure  ;

increase. The resulting internal pipe expansion would aid in reducing the internal (

f pipe pressure, possibly below code allowable and ultimate strengths.

c. Conclusions The inspector concluded the lack of overpressure relief protection on the PCC i system was a design defect that had existed since the start of commercial plant  !

operation. This design deficiency was apparently overlooked and represented ,

. misinterpretation (at the time) of the design requirements of Standard B31.1.. This i deficiency was not discovered until Maine Yankee had commenced a' review of the  !

PCC system in July 1996, before the arrival of the NRC Independent Safety . {

Analysis Team (ISAT). Once Maine Yankee and YAEC discovered this issue, j appropriate measures were taken to correct the design deficiency.- The NRC has - l since issued Generic Letter 96-06 to the industry.

Notwithstanding Maine Yankee's corrective actions, Technical Specification (TS) 3.6, " Emergency Core Cooling and Containment Spray Systems," and Technical Specification 3.12,." Station Service Power," require, in part two operable i trains of ECCS equipment and two operable EDGs whenever the reactor coolant i system temperature is greater than 210*F or 400 psig. The lack of relief valve f

protection on the PCC system had the potential to render components cooled by the PCC system (i.e., "A" train ECCS and one of the emergency diesels) inoperable (apparent violation of plant YS 3.6 and 3.12).  !

E3 Engineering Procedures and Documentation E3.1 Desian and i.icensina Bases Discreoancies (URI 50-309/96-16-08, throuah l 50-309/96-16-15 l

a. Insoection Scone (37551,92903) j As a result of the independent Safety Assessment (ISA) team's reviews that were '

conducted to determine Maine Yankee's conformance to its design and licensing basis, a number of issues and concerns were identified. Those issues were further-  !

reviewed by NRC Region I concerns consistent with the overall purpose of this f inspection.

b. Observations. Findinas and Conclusions

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in the area of transient and accident safety analyses, the following issues were identified: ,

(1) Safety Evaluation Report (SER) Conditions Satisfied but not Documented. 3 f

!n Section 2.1.2, of the ISA report, the tsam determined that Yankee Atomic l Electric Company (YAEC) did not have a written process to document how safety }

analyses conformed to SER conditions. While conditions of use specified in the  :

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SERs were found to be satisfied, they were not documented. In response to this concern, MY initiated actions to develop and implement a process for documenting SER compliance for all safety analysis methodologies and codes. This item remains open pending completion of NRC's reviews of the following of Maine Yankee's ;

process for documenting SER compliance for all safety analyses methodologies and codes. (URI 50-309/90-16-08).

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(2) Main Steam Line (MSL) Ruoture Analysis Errors and Inconsistencies:

l In Section 2.1.4 of the ISA Report, a number of errors and inconsistencies were documented by the team. Maine Yankee's actions in this area has not been ,

reviewed by the NRC. This item remains open pending completion of NRC's reviews of Maine Yankee's MSL rupture analyses errors and inconsistencies.

(URI 50-309/96-16-09).  ;

r (3) Lack of a Documented Process to Demonstrate Code Capability in Section 2.1.5, the ISA identified that YAEC did not have a documented process to demonstrate code capability. MY has initiated corrective actions with YAEC to write a Methods Overview Manual to provide a document process. This item remains open pending completion of NRC's reviews of Maine Yankee's process for ,

demonstrating code capability. (URI 50-309/96-16-10).  ;

in the area of design review of selected systems, the following issues were 7 identified: i (4) Containment Sorav (CS) System and the Component Coolina Water (CCW)

Systems As documented in ISA Report Sections 2.2.1.4 and 2.2.2.6, The ISA team identified !

a number of concerns involving safety system analyses. Specifically, the capability ;

of the CS and CCW systems to meet the design basis assumptions for a Loss of l Coolant Accident (LOCA) initiated from greater than 2440 MWt was questioned.

This item remains open pending completion of NRC's review of the CS and CCW system capability to meet the design basis assumptions for a LOCA initiated from greater than 2440 Mwt. (URI 50-309/96-16-11).

A second issue identified in Section of 2.2.2.6 of the ISA Report, involved the development of a program for CCW and Residual Heat Removal (RHR) heat l exchanger fouling and testing. As a result of ISA team review in this area, the basis ;

for the value for the remaining loads (i.e., allloads after the maximum containment t sump temperature) on the CCW heat load was questioned. This item remains open :

pending completion of NRC's reviews of the CCW and RHR heat exchanger fouling !

and testing methodologies and the basis for the value of the CCW heat loads.

(URI 50-309/96-16-12).

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(5) Residual Heat Removal (RHR) Heat Exchanaer Thermal Transient

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i in Section 2.2.2.5 of the ISA Report, the team found that with the plant operating 3 at 2700 MWt, the most limiting design-basis thermal transient, shifting to recirculation during a LOCA, on the RHR heat exchangers was beyond their design ,

specification and had not been analyzed. This item remains open pending ,

completion of NRC's review of the adequacy of the RHR Hx in the recirculation  !

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mode during a LOCA from 2700 Mwt. (URI 50-309/96-16-13).

l (6) Electrical Calculations For Emeraency Diesel Generator (EDG) l l

In Section 2.3.6 of the ISA Report, a number of electrical calculational problems l were noted for the EDGs. These problems involved the use of incorrect motor data and cable data, as well as a problem with breaker coordination curves reflecting omitted information for cable data and cable damage curves. This item remains open pending completion of NRC's review the adequacy of the electrical j calculations. (URI 50-309/96-14).

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(7) Environmental Qualification (EO) Issues Sections 2.3.9.2 of the ISA report identifies potentially significant EQ issues that j remain open. One issue involves the assumptions used to establish EQ j requirements for Emergency Feedwater (EFW) were inconsistent with Emergency l Operating Procedure E-1, Loss of Primary or Secondary Coolant. The second area for potential EQ issues involves the recently expanded high energy line break concerns in the turbine building, notably the battery and protected switchgear ,

rooms. Section 2.3.9.3 identified a substantial backlog of EQ work. ISA Report !

Section 2.3.7.1, which documents ISA team identified additional vulnerabilities that I could have resulted in a harsh environment for safety-related electrical equipment-located in these areas. This item is an unresolved pending further NRC staff review.

(URI 50-309-96-15).

E3.2 FSAR Discrepancies (URI 50-309/96-16-16 throuch 50-309/96-16-24)

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- a. insoection Scope (37551. 92903)

As a result of the ISA team review of MY operations, a number of issues concerning conformance with the Updated Final Safety Analysis Report (UFSAR) were identified. These issues were further reviewed by the NRC Region I and classified as follows:

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b. Observations. Findinos and Conclusions  ;

I (1) ECCS System Conformance to FSAR Criterion 44 l

In Section 2.2.1.4 the ISA Team noted concerns about the ability of the CS system to provide a reliable supply of water during the recirculation phase of a LOCA. As documented in section E3.1, further NRC review of this matter is planned. The ISA team further noted that their review of current and historical design calculations to

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evaluate if NPSH available for CS pumps during the recirculation phase was derived !

with appropriate conservatism required by UFSAR Criterion 44. This Criterion _

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provided requirements for ECCS and stated in part, "The performance of each ECCS

shall be evaluated conservatively in each area of uncertainty." This item remains

! open pending completion of NRC's' review of MY's conformance to UFSAR Criterion (

44. (URI 50-309/96-16-16). i i

(2) Operation of the Comoonent Coolina Water System Section 2.2.2.6 of the ISA Report documented that based upon MY 's analysis, the !

CCW systems would not support plant operation up to the SW temperature values l in the UFSAR. Specifically, UFSAR Section 9.4.1 stated, "The component cooling l water system heat balance was performed la 1990... demonstrating adequate i capacity for design basis post-LOCA conditions assuming a service water (SW) inlet l temperature of 80 degrees F for CCW heat exchangers E-4B and E-5A, and 90 degrees F for CCW heat exchangers E-4A and E-5B." However, an engineering directive that was based upon an integrated heat removal analysis provided for ,

operating limitations that were not bounded by the aforementioned licensing l'

commitments (i.e., the FSAR). The engineering directive restricted the maximum -

SW temperature to 70.2 degrees F for E-4B and E-5A, and 78.5 degrees F for E-4A i and E-5B. This inability of the SW system to support plant operations up to the i temperature values in the FSAR without an appropriate change in place appears to be a violation of 10 CFR 50.59. (Apparent Violation of 10 CFR 50.59).  :

.i (3) Reoortability of CCW Operations Different Than FSAR Desian Flow tests were performed by MY during the 1995 refueling outage to verify that )

the flow distributions in the PCC and SCC systems satisfy design basis  !

requirements. The results of these flow tests were evaluated and compared to I design hydraulic calculations of the PCC/ SCC systems in Technical Evaluation (TE) i No.121-96, dated November 22,1996. As part of this TE, flow diversion as caused by the leakage through 10" PCC-T-20 has been evaluated to support a 3

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reportability determination. This issue was associated with a 1995 event involving the discovery that the CCW heat exchanger bypass valve, PCC-T-20, was open about 11 degrees had not had a reportability determination completed at the time of the ISA review. Subsequently on December 2,1996, MY completed a reportability determination. The basis that event was not reportable was based upon the results of the TE.

The TE also contained a 10 CFR 50.59 evaluation of the :.ubject flow testing. The issue of the UFSAR wording in Section 9.4.1 was addressed and proposed new wording was provided that would remove a limitation on maximum service water inlet temperature to the CCW heat exchangers.

The acceptability of MY conducting operations with CCW systems that would not support plant operations up to the SW temperature values in the UFSAR warrants further NRC review. This review would address the acceptability of the reportability l

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determination made by MY for the PCC bypass valve mispositioning. This item is considered open pending completion of the NRC staff review. (URI 50-309/

96-16-17).

(4) 115kV Offsite Power Lines ISA report, section 2.3.1,115 kV Offsite Power Lines, documents the team's conclusion that the limited capability of the Suroweic line was contrary to the design and licensing basis presented in UFSAR Section 8.2.3, which stated that either of the 115 kV lines was independently capable of supplying the plant auxiliary power system. The ISA team did not consider MY's position that the 345 kV system back-feed operation, completed within six hours, was an acceptable basis for compliance with the UFSAR and MY Design Criterion 39 stated in Appendix A of the UFSAR.

In a November 21,1996 letter to MY, NRC:NRR requested that MY respond to question that, in part, were intended to clarify the licensing basis of the plant associated with the 115 kV lines and measures to resolve this issue. This matter remains an unresolved pending further MY actions and the NRC:NRR review.

(URI 50-309/96-16-18).

(5) Dearaded Grid Undervoltaae Relav Calibration Tolerance Band in Section 2.3.2 of the ISA Report, the team found that the calibration tolerance band for the degraded grid undervoltage relays may result in a setting that could cause a premature transfer of loads from offsite power to the Emergency Diesel Generators (EDG) following a LOCA. This could have created a situation that would be contrary to UFSAR Section 8.2.3, which required the offsite reserve power system to be capable of supplying the plant auxiliary power system. This matter warrants further NRC review and is considered open. (URI 50 309/96-16-19)

(6) Emeraency Diesel Generator Electrical Loadina Section 2.3.5 of the ISA Report identified a condition involving MY's assessment of

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EDG loading to be non-conservative. This condition included concerns involving:

(1) all required loads were not included in the subject calculation; (2) the loading profile was inconsistent with the UFSAR and design requirements; and (3) a weakness in the area of configuration control involving the replacement of MOV operators due to the failure to account for motor horsepower changes as required by procedure 17-227, Electrical Distribution System Load Tracking. The issues of conformance to UFSAR requirements, and design and configuration management controls remains open pending completion of further NRC review. (URI 50-309/

96-16-20).

(7) Protected Switchaear Room Ventilation A vulnerability in the support of compensatory operator actions specified in the UFSAR for portable ventilation to be set up in the protected switchgear room was identified by the ISA team in Section 2.3.7.1 of the ISA Report. Specifically, in UFSAR Section 8.3.3, MY described the lack of redundancy for this ventilation

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system (single supply and single exhaust fans) and specified operator actions required to mitigate the consequences of a single fan failure. The review by the ISA team identified that no emergency power source was available to power portable fans during a design-basis event which may include loss of normal ac power.

Therefore, the compensatory operator actions to open doors and set up portable fans described in the UFSAR and MY's analyses were not technically supported.

This item remains open pending completion of NRC's review of the potential impact of this vulnerability on plant operations. (URI 309/96-16-21).

(8) Procedure for Cross-Connectina DC Buses in ISA Report Section 2.3.8 identified that procedure 1-22-2, AC and DC Vital Bus Operation, allowed cross connecting redundant 125 Vdc vital buses for up to 72-brs during plant operation, contrary to FSAR Appendix A, Criterion 39, Emergency Power for ESFs. This criterion provides, in part, that the alternate power systems shall be provided and designed with adequate inebpendency and redundancy to permit the functioning required of the ESFs and n a minimum the onsite power system shallindependently provide required capacity assuming a single failure. The absence of an adequate 10 CFR 50.59 review allowing this condition to exist appears to be in violation of 10 CFR 50.59 (Apparent Violation of 10 CFR 50.59).

(9) FSAR Inconsistencies Section 2.4 of the ISA team report documented that over 100 issues were identified by MY prior to ISA team site activities, and which occurred as a resmt of an i initiative to upgrade the UFSAR. Subsequently, the ISA team and NRC Region I ,

inspections identified additional FSAR and 50.59 related inconsistencies, concerns, j or deficiencies. The ISA report also noted in Section 3.2.4 that despite many

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inconsistencies between information contained in the FSAR and the as-built )

condition of the plant, MY did not recognize a need to review and reconcile test j procedures with design requirements and to correct these errors. These testing '

inadequacies are discussed in Section M1.2.

An assessment of licensee identified and some ISA team identified conditions was performed by MY that identified: (1) approximately 89 of these changes needed a 10 CFR 50.59 evaluation or reviews to support the corrective text reflecting equipment and procedures that have changed from that described in the UFSAR; and (2) there were 27 changes to the UFSAR that should have been made as a result of implementation of either Engineering Design Change Requests or Plant Design Change Requests that have already been implemented (15 of these were dated from 1989 to 1995) at the facility. Also, this MY assessment reviewed the UFSAR upgrade packages for operability /reportability considerations, of which no deficiencies were identified.

I While the follow up inspection conducted in this inspection period did not identify specific MY administrative requirements pertaining to the performance of 50.59 l

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Safety Evaluations or periodic UFSAR updates, the above enumerated conditions are considered by the NRC to be instances of apparent violations of the requirements of:

10 CFR 50.59(b)(1) to provide written b'ases that changes that are made to the facility or its operation do not result in Unreviewed Safety Questions; and that annual updates to the UFSAR are to be made in accordance, with 10 CFR 50.71(e)(4) (Apparent Violations of 10 CFR 50.59(b)(1) and 10 CFR 50.71(e)(4) ).

(10) Other FSAR inconsistencies Also in this section of the ISA Report, two other inconsistencies with the UFSAR were addressed by the ISI. These were: l i

e Spent Fuel Pool (SFP) Heat Exchanger Rating  ;

I The ISA team's preliminary review documented in Section 2.4.1 of the ISA j I- Report of the SFP heat exchanger discrepancy in Section 9.8.2 (i.e., j s nameplate rating of 200 degrees F vs. the 225 degree F UFSAR value) '

, concluded that MY's plans to further analyze this condition by January 1, j 1996 was appropriate. This issue is considered unresolved pending the i completion of MY's analysis and the NRC's review of that analysis j

,

(URI 50-309/96-16 22).

j e Atmospheric Steam Dump (ASD) rated for 2.5 Percent Power

~

in Section 2.4.2 of the ISA Report, the team determined that assumptions made regarding the response to a steam generator tube rupture (SGTR) were not realistic. This section also documented the fact that MY assumed that the capability of the ASD was 5 percent power as part of its validation for EOPs.

In July 1996, MY informed the NRC that the ASD valve's capability was 2.5

, percent power. While UFSAR section 14.12 stated that the quantity of

- reactor coolant transported through the leak to the steam system is the same with or without offsite power, the ISA team was concerned that both the

, loss of offsite power and the reduced capability of the ASD would result in the time needed to isolate the affected steam generator to be longer than that stated in the UFSAR.

Both the acceptability of the EOPs and the reduced capability of the ASD valve as they relate to the SGTR event has been the subject of on-going NRC review. Also, prior to compensatory measures being put in place (i.e., providing additional operator training and establishing an administrative limit on reactor coolant system activity), the potential existed to overfill the steam generators thereby releasing radioactive liquids to the environment, and possibly exceeding the limits of 10 CFR Part 100. This item remains open pending completion of NRC's review of the

.

effects of the reduced capacity of the ASD on a SGTR event. (URI 50-309/

.

96 18-23).

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(11) 50.59 Safety Evaluation inadeausev for Control Room Ventilation Testina  !

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Section 3.2.4, item (2) of the ISA report identified a failure to properly evaluate the -

,

October 31,1995 control room ventilation test results. This item involved the.

.

adequacy of a surveillance test used to determine that the control room is  !

'

! maintained at a positive pressure following accident conditions, and a subsequent j. operability determination by MY that a negative pressure corsdition was acceptable. +

- The "A" train of the control room breathing air supply (CRBAS) system failed to l i

provide for the demonstration of a positive pressure, until the surveillance test was .

performed satisfactorily on August 17,1996

*

! Surveillance testing of the control room ventilation system (a two train system made l up of a recirculation system and a CRBAS is performed in accordance with >

procedure 3.17.5, Surveillance testing of Tech. Spec Charcoal and HEPA Filters. ,

Procedure Step 5.3.3, requires the verification of a positive pressure within the

, control room during testing of the CRBAS system. Technical Specification (TS)

'  ;

Section 3.25.B, requires that two trains of control room ventilation shall be operab?e i

whenever the reactor is critical. Technical Specification Section 4.11.B.2.a i

! specifies that filter system operability of the control room recirculation and CRBAS )

! system is demonstrated at least each refueling interval, and TS Section 4.11.B.3.d  ;

} specifies that filter system operability is determined by verifying that the CRBAS .

system flow rate is at least 40 cfm during system operation. The Basis for TS l Section 3.25.B states that: (1) the operability of the control room ventilation system I'

ensures that the control room will remain habitable for operations personnel during

and following all credible accident conditions; (2) that the operability of this system i- in conjunction with control room design provisions is based on limiting the radiation i exposure to personnel occupying the control room to 5 rem or less whole body, or i its equivalent; and (3) this limitation is consistent with the requirements of General

! Design Criteria 19 of Appendix A,10 CFR 50. ,

i

There appears to be no explicit TS requirement pertaining to the CRBAS system maintaining the control room in a positive pressure condition. However, the NRC

[ considers that the surveillance procedure 3.17.5 requirement for the CRBAS system i to demonstrate a positive pressure in the control room, coupled with fact that control room operation at a slight positive pressure in the accident mitigation

<

function is specified in the MY Control Room Ventilation System Design Bases Summary Document, to imply that the demonstration of each train of the CRBAS j system to provide for the development of a positive control room pressure to be a j'

condition of operability. Based on the aforementioned requirements, the failure of MY to demonstrate that two trains of control room ventilation were operable prior to 1 making the reactor criticalin January,1996 was an apparent violation of TS , 3.25.B.2 (Apparent Violation of TS 3.26.B.2) )

.

In response to the knowledge of the unsatisfactory surveillance result involving the  ;

E

"A" CRBAS train, MY requested that Yankee Nuclear Services Division developed

control room radiological analysis. The provided analysis, dated December 12,

1995, incorrectly determined that the subject failed surveillance test had l demonstrated a positive pressure. Based upon MY engineering personnel use of this

,

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analysis to support the developmerit of a 10 CFR 50.59 safety evaluation that provided that the negative control raom pressure condition was not an Unreviewed Safety Question (USQ) and believ'ng that the analysis provided acceptable margins to safety with respect to Critenon 19 of Appendix A to 10 CFR 50, an inadequate 50.59(b)(1) analysis was performed. The failure to provide an acceptable safety evaluation that provides the written basis that operation of the plant without both trains of the CRBAS system demonstrating the capability to provide for positive i control room pressurization is not a USQ is considered an apparent violation of l 10 CFR 50.59(b)(1) (Apparent Violation of 10 CFR 50.59(b)(1)).

It was noted that ventilation system lineups (i.e., ventilation fans running that would not have power in design basis accident scenario) could create conditions were the control room might not maintain a positive pressure with respect to an adjacent area. As such, this condition may result in a more limiting case and needs actions for operator actions to ensure planned response to accident conditions area properly l bounded by existing analysis. The inspector was unable to identify specific ]

instruction used by MY to establish test lineup conditions with respect to ventilation I system lineups. This item remains open pending completion of NRC's review. !

(URI 50-309/96-16-24). l E3.3 Desian/Encineerina Related issue (URI 50-309/96-16-25 throuah 27) l

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a. Inspection Scoce (37551,92903) j in sections 3.3.2.2 and 3.3.3 of the ISA report, there were some weaknesses noted in engineering programs.

b. Observations. ?indinas and Conclusions i Weaknesses were noted in the Erosion Corrosion (E/C) Program because Maine l Yankee had no firm commitment to use a specific code such as EPRI CHECWORKS. '

A review of the program revealed weaknesses such as ome inspected fittings ;

subsequently leaked and a component identified with E/C was not replaced. This i program requires further NRC review and as such, this item remains open pending completion of the review. (URI 50-309/96-16-25).

In Section 3.3.3 of the report, it was noted that the quality and availability of Design Basis Information (DBI) varied. Design basis summary documents (DBSDs) could not be completely relied upon due to discrepancies. MY had identified weaknesses in the inputs and Assumptions Source Document (IASD) and this was being replaced by more comprehensive Safety Analysis information Document (SAID).

Subsequent NRC review of the SAID document is needed and this item remains open pending completion of the review. (URI 50-309/96-16-26) i Some weaknesses were also noted in the control of calculations as discussed in section 3.3.4 of the ISA Report. There were two examples of conflicting calculations cited. These weaknesses will further be reviewed by the NRC and as such this item romains open. (50-309/96-16-27).

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E8 Miscellaneous Engineering issues

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E8.1 Corrective Actions (URI 50-309/96-16-28 and 29)

a. Insoection Scope (40500)

b. Observations. Findinas and Conclusions The ISA team had identified a number of significant weakness and deficiencies which were related, in part, to a root cause involving poor problem identification.

,

Furthermore, the ISA noted the existence of certain design errors, in some cases I MY was aware of them and had failed to take action to address them. A number of issues were reviewed in this and other inspections associated with ISA issue follow-

"

up that relate to MY's failure to identify or promptly correct significant problems.

The items enumerated below represent areas that were conditions adverse to quality 4 existed and where MY failed to recognize their significance or take timely corrective action.

(1) Section 4.2 of the ISA report concluded that overall, MY's Corrective Action Program (CAP) was weak and resulted in instances of untimely and ineffective corrective actions. The weakness in the CAP is considered an

-

Unresolved Item which will receive further review during an upcoming NRC inspection (Inspection Procedure 40500). (URI 50-309/96-16-28).

(2) Design Deficiency issues that represented instances of inadequate corrective actions and are addressed in other sections of this report are:

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(1) Ventilation Issues: ISA Report Sections 2.3.7.2, 3.3.1, and 4.2.2: In 1991 a YNSD Engineer identified the potential to lose safety related ventilation flow into the CS building due to the closure of ventilation dampers utilizing a non safety-related source of instrument air. The ISA team considered MY's performance weak in regard to the lack of follow up of the identification of damper concerns from 1991 and also because the recent focus on the problems with the CS building ventilation system did not result in identifying a vulnerability with the dampers. Specifically, several efforts had been extended at dealing with the icing and clogging of the CS building HVAC unit, HV-7.

(11) Atmospheric Steam Dumo Valve: In 1986, MY identified that the atmospheric steam dump (ASD) valve will only relieve 2.5 percent power. A 5 percent

, power capacity is required to support EOPs used to respond to inadequate core cooling events. Plans to upgrade ASD capacity was dropped as an issue requiring resolution. This issue was resurfaced by MY in March 1996, and the NRC was informed about the condition by MY in accordance with 10 CFR 50.9. As documented in ISA report Sections 3.3.1 and 4.2.2, this issue was an example of a deficient design condition that had not been entered into a corrective action system and was dropped. Further NRC review of this isssue is considered an unresolved item. (URI 50-309/96-16-29).

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(Ill) Environmental Qualification of Comoonents in Containment: This issue is discussed in detail in Inspection Report 96-10, and involves the ISA team's

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identification of components in the containment that were below submergence levelin the EQ Program. This issue was considered by the ISA in report Section 4.1 to be an example of a significant condition adverse to quality that went unrecognized by MY.

(IV) Turbine Buildino Floodino: Section 4.2.2 of the ISA report documents that the

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Service Water System Operational Performance inspection performed by MY in 1994 identified an item involving turbine hall flooding which was scheduled to be addressed in June 1997. However, the April 1996 IPEEE findings showed that the plant was outside of its design basis for a turbine hall flood. This design issue could have been identified and resolved in 1994.

This item is a corrective action item that was determined by the ISA team to represent an untimely correction of a significant issue.  !

(3) Lack of Complete Testing of the ECCS actuation logic. As documented in i ISA report Section 4.1, this area represents another example of a significant condition adverse to quality that went unrecognized by MY. The NRC has !

identified in inspection Report 96-11 apparent violations of TS requirements '

as a result of MY's failure to conduct required testing.

(4) Section 4.2.2 of the ISA Report identified two instances of MY's corrective

,

actions that had been ineffective, and in which one of the instances caused repetitive problems. An example of the latter case was the AFW control

, system which has been the subject of repetitive problems. The manner in which MY addressed the negative control room pressure identified during a surveillance conducted in October 1995 was an example of ineffective corrective actions.

The issues addressed in items (2)(1), (2)(111), (2)(IV), (3) and (4) above,

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represent instances of either conditions adverse to quality that were not identified, or not promptly corrected, or the corrective actions did not prevent recurrence and therefore were considered failures to correct conditions adverse to quality as required by 10 CFR 50, Appendix B, Criterion XVI.

(Apparent Violation of 10 CFR 50, Appendix B, Criterion XVI).

(5) As documented in Section 4.3.3 of the ISA Report, MY deferred the specialty training program for mechanical, electrical and I&C personnel in order to increase the availability of maintenance workers to support the 1995 outage work. Also, due to resource constraints, the MY training organization was unable to respond to industry reaccreditction issues. The effects of these deferrals are to be further reviewed by the NRC and as such, this item is considered unresolved. (URI 50-309/96-16-30).

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IV. Plant Suooort R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 Radiation Protection Proaram (URI 50-309/96-16-30)

a. Inspection Scoce (92904)

b. Observationt. Findinas and Conclusions The ISA team identified in Section 3.1.4.3 of their report that approximately 50 percent of personnel contamination events identified by MY in 1996 occurred in areas that were believed to be clean or uncontaminated. This matter warrants further NRC review and is considered an open pending completion of the review.

(URI 50-309/96-16-31).

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V. Manaaement Meetinas X1 Exit Meeting Summary The inspectors presented the inspection results to members of tim licensee on January 28, 1997. The licensee acknowledged the findings presented, i l

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. PARTIAL LIST OF PERSONS CONTACTED

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Licensee

+ P. Anderson, Yankee Nuclear Services Division.....(Via Telecon)  ;

+W. Ball, Assistant Manager, Operation Department

+W. Barry, Section Head, Plant Engineering Department i

) E. Brand, Nuclear Safety Engineering Group

+W. Baxter, Plant Shift Superintendent

+ R. Blackmore, Plant Manager i

+J. Connell, Manager, Technical Support Department

+C. Frizzle, President and Chief Executive Officer.....(Via Telecon) *

+ R. Hayward, Manager, Training Department

+J. Hebert, Manager, Licensing and Engineering Support

+S. LeClerc, Section Head, Quality Programs Department-j + G.' Leitch, Vice President - Operations  :

- M. Marston, Section Head - Plant Engineering ,

+J. McCann, Section Head, Licensing

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+ R. Meixell, Section Head, Maintenance Department

+ S. Nichols, Manager, Corporate Engineering Department

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+J. Niles, Assistant Manager, Operations Department 1

. + E. Soule, Manager, Plant Engineering Department l 1 + S. Smith, Manager, Operations Department 1

+J. Weast, Licensing Engineer j. + M. Whitney, Licensing Engineer...........(Via Telecon)

[ + D. Whittier, Vice President - Engineering......(Via Telecon)

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+ R. Conte, Chief, Reactor Projects Branch 5 D. Dorman, NRR Project Manager H. Einchenholz, Project Engineer K. Kolaczyk, Reactor Engineer

+ R. Rasmussen, Resident Inspector W. Olsen, Resident inspector

+J. Yerokun, Senior Resident inspector Other

+P. Dostie, State of Maine Nuclear Safety ins,pector  !

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+ Denotes those present at the exit meeting held on January 28,1997.

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lNSPECTION PROCEDURES USED ,

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IP 37551: Onsite Engineering >

IP 40500: Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Prcblems

IP 61726: Surveillance Observation IP 62707: Maintenance Observation

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IP 71707: Plant Operations IP 92901: Follow up - Operations  ;

lP 92902: Follow up - Maintenance IP 92903: Follow up - Engineering IP 92904: Follow up - Plant Support

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ITEMS OPENED, CLOSED, AND DISCUSSED

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Items Opened:

50-309/96-16-01 URI Further review required to determine the appropriateness of some Maine Yankee Technical Specifications Interpretation required. (O4.4)

50-309/96-16-02 URI lssue of unavailability of the plant process computer during plant transients and the adequacy of the post trip reviews is considered unresolved pending further NRC review. (04.5)

50-309/96-16-03 URI Further NRC review of the isolation time for a SGTR to ascertain the adequacy of the performance of 50.59 reviews associated with the establishment of more restrictive TS limits on reactor coolant activity is needed. (O4.6)

50-309/96-16-04 URI Further review of the testing and the reset of the HPSI throttle valves during the next system testing is needed. (M1.2b(1))

50-309/96-16-05 URI The calibration and periodic testing requirements of standby Power Meters Required by R.G.1.97 needs to be further reviewed. (M1.2b(2)(VI))

50-309/96-16-06 URI The adequacy of MY IST Program and/or 10 CFR 50 Appendix B Criterion XI program requires further reviews. (M1.2b(2)(VI))

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27-50-309/96-16-07 URI Further review of the as-found conditions in the SW pump bay, and the adequacy of MY's inspection of the containment at the completion of an outage is required. (M2.1)  ;

50-309/96-16-08 URI Further NRC's reviews of Maine Yankee's process for ;

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documenting SER compliance for all safety analyses methodologies and codes; process for demonstrating code e capability and the MSL rupture analyses errors and

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inconsistencies are needed. (E3.1b(1))

50-309/96-16-09 URI Further NRC review of Maine Yankee's MSL rupture analyses ;

errors and inconsistencies is required. (E3.1b(2)) ;

50-309/96-16-10- URI Further NRC review of Maine Yankee's process for demonstrating code capability is required (E3.1b(3))

l 50-309/96-16-11 URI Further NRC review of the CS and CCW system capability to {

meet the design basis assumptions for a LOCA initiated from greater than 2440 Mwt is required (E3.1b(4))

50-309/96-16-12 URI Further NRC review of the CCW and RHR heat exchanger fouling and testing methodologies and the basis for the value of the CCW heat loads is required (E3.1b(4))

50-309/96-16-13 URI The adequacy of the RHR Hx in the recirculation mode during a LOCA from 2700 Mwt needs further NRC reviews. (E3.1b(5))

50-309/96-16-14 URI The adequacy of the electricalload calculations for the emergency diesel generators need further reviews. (E3.1b(6))

50-309/96-16-15 URI The substantial backlog of EQ work, and the potential vulnerabilities that could have resulted in a harsh environment for safety-related electrical equipment located in certain areas need further NRC review. (E3.1b(7))

50-309/96-16-16 URI The evaluation of the performance of each ECCS to conform to UFSAR Criterion 44 needs further reviews. (E3.2b(1))

50-309/96-16-17 URI The acceptability of the reportability determination made by MY for the PCC bypass valve mispositioning requires further NRC j review. (E3.2b(3))

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50-309/96-16-18 URI Further reviews are needed to clarify the licensing basis of the plant associated with the 115 kV lines and measures to resolve ;

the inadequacies if any. (E3.2b(4)) l l

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50-309/96-16-19 URI The potential for the calibration tolerance band for the degraded grid undervoltage relays resulting in a setting that could cause a premature transfer of loads from offsite power to the Emergency Diesel Generators (EDG) following a LOCA, thereby creating a situation that would be contrary to FSAR Section 8.2.3 requires further reviews. (E3.2b(5))

50-309/96-16-20 URI The issues of conformance to UFSAR requirements, and design and configuration management controls remains open pending completion of further NRC review. (E3.2b(6))

50-309/96-16-21 URI Compensatory operator actions to open doors and set up !

portable fans described in the UFSAR and MY's analyses were '

not technically supported. This item remains open pending completion of NRC's review of the potential impact of this vulnerability on plant operations. (E3.2b(7))  !

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l 50-309/96-16-22 URI The discrepancy with the spent fuel pool heat exchanger j (nameplate rating of 200 degrees F vs. the 225 degree F '

UFSAR value) requires further analysis by MY and the NRC's review of that analysis. (E3.2b(10))

l 50-309/96-16-23 URI Further review of the effects of the reduced capacity of the 1 ASD on a SGTR event is required. (E3.2b(10))

50-309/96-16-24 URI NRC review of the specific instruction used by MY to establish test lineup conditions with respect to ventilation system lineups is needed. -(E3.2b(11))

50-309/96-16-25 URI Further reviews of the weaknesses noted in the Erosion Corrosion (E/C) Program are needed. (E3.3)

50-309/96-16-26 URI NRC review of the Safety Analysis Information Document is required. (E3.3)

50-309/96-16-27 URI Further reviews of the weaknesses noted in the control of calculations are needed. (E3.3) ,

J 50-309/J6-16-28 URI Weakness in the Corrective Action Processes need further NRC i

review during the upcoming 40500 inspection (E8.1b(1))

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50-309/96 16-29 URI Further NRC review of the issue of the adequacy of Maine j Yankee's use of the Corrective Action System as applied to the ;

atmospheric steam dump's capacity is needed. (E8.1b(2)(ll))

50 309/96-16-30 URI The effects of MY deferral of the specialty training program for l mechanical, electrical and l&C personnel in order to increase the j availability of maintenance workers to support the 1995 outage work needs further review by the NRC. (E8.1b(5))

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.50-309/96-16-31 URI Further NRC review of the personnel contamination events -[

identified by MY in 1996 is needed. (R1.1)

Items Closed: None ,

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LIST OF ACRONYMS USED AEOD Office for Analysis and Evaluation of Operational Data ,

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AFW Auxiliary Feedwater *

ALARA As Low As is Reasonably Achievable AOP Abnormal Operating Procedure ASDV Atmospheric Steam Dump Valve '

CCW Component Cooling Water CFR Code of Federal Regulations CIS Containment isolation System CRBAS Control Room Breathing Air Supply CSAS Containment Spray Actuation System CVCS Chemical and Volume Control System DRP Division of Reactor Projects DRS Division of Reactor Safety EA Escalated Action ECCS Emergency Core Cooling System EDG Emergency Diesel Generator EFW Emergency Feedwater EP Emergency Preparedness EQ Environmental Qualification ESF Engineered Safeguards Feature FR Federal Registry FSAR Final Safety Analysis Report gpm gallons per minute HPSI High Pressure Safety injection IFl Inspection Follow-Up Item r IFS Inspection Follow-Up System ISA Independent Safety Assessment lMC Inspection Manual Chapter IPEEE Individual Plant Evaluation for External Events ,

ISI in-Service Inspection IST In-Service Testing LER Licensee Eve LOCA Loss of Coolant Accident LPSI Low Pressure Safety injection MOV Motor Operated Valve NCV Non-Cited Violation NOV Notice of Violation NPSH Net Positive Suction Head '

NRC Nuclear Regulatory Commission NRR Nuclear Reactor Regulation (office of) ,

OE Office of Enforcement PCC Primary Component Cooling PTR Post Trip Review QA Quality Assurance

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RA Regional Administrator RAS Recirculation Action System

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31 i l

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RHR Residual Heat Removai RP Radiation Protection l RP&C Radiological Protection and Chemistry RWST Refueling Water Storage Tank SALP Systematic Assessment of Licensee Performance SCC Secondary Component Cooling SER Safety Evaluation Report SFP Spent Fuel Pool SGTR Steam Generator Tube Rupture SI Safety injection SIAS Safety injection Actuation System TE Technical Evaluation Tl Temporary Instruction TS Technical Specification UFSAR Updated Final Safety Analysis Report UOR Unusual Occurrence Report URI Unresolved item USQ Unreviewed Safety Question WO Work Order YAEC Yankee Atomic Energy Company YNSD Yankee Nuclear Services Division i