IR 05000295/1987028

Insp Repts 50-295/87-28 & 50-304/87-29 on 870513-0821.No Violations or Deviations Noted.Major Areas Inspected:Actions on Previous Insp Findings,Implementation of TMI Action Items & Emergency Operating Procedures

ML20235V723
Person / Time
Site:	Zion  File:ZionSolutions icon.png
Issue date:	10/09/1987
From:	Hinds J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20235V709	List: IR 05000295/1987028 ML20235V706
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-1.C.1, TASK-1.D.1, TASK-1.D.2, TASK-2.B.1, TASK-2.B.2, TASK-2.E.1.2, TASK-2.E.4.2, TASK-2.F.1, TASK-2.K.3.01, TASK-2.K.3.05, TASK-2.K.3.25, TASK-3.D.3.4, TASK-TM 50-295-87-28, 50-304-87-29, NUDOCS 8710150294
Download: ML20235V723 (41)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Report Nos. 50-295/87028(DRP);50-304/87029(DRP)

Docket Nos. 50-295; 50-304 License Nos. DPR-39; DPR-48 Licensee: Commonwealth Edison Company P. O. Box 767 Chicago, IL 60690 Facility Name: Zion Nuclear Power Station, Units 1 and 2

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Inspection At: Zion, IL Inspection Conducted: May 13 through August 21, 1987

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Inspector: Donald A. Beckman Prisuta-Beckman Associates

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Approved By J. M. Hinds, J ef io.o9 67 eactor Projects Section 1A Date Inspection Summary Inspection on May 13 through August 21, 1987 (Report Nos. 50-295/87028(DRP);

50-304/87029(DRP))

Areas Inspected: Routine, announced inspection of licensee actions on previous I inspection findings; licensee implementation of Three Mile Island (TMI) Action Plan items in accordance with NUREG 0737, " Clarification of TMI Action Plan Requirements"; and emergency operating procedure Results: No violations or deviations were identifie Ten TMI Action Plan items were closed as acceptable. One potential deviation was identified (see paragraph 3.e).

8710150294 G71009 PDR ADOCK 05000295 0 PDR

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DETAILS

Persons Contacted

• G. Plimli' Station Manager-L*T. Rieck, Superintendent, Services

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• R. Budowle, Assistant Station Superintendent, Technical Services

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• N. Valos, Unit 2 Operating Engineer

• C, Schultz, Regulatory Assurance Administrator

• J. Ballard, Quality Control Supervisor

• L.' Obremt, Quality Assurance Engineer

• A. Bless, Regulatory-Assurance Engineer i J. Carlson,' Project . Engineer

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D. Clark, Nuclear Training Supervisor

• D. Dahlen,LTechnical Staff Group Leader S Kapsalis, Technical Staff Engineer G. Kassner, Health. Physicist

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• L. LaSpisa, Assistant Technical Staff Supervisor P. LaBlond, Nuclear Licensing Assistant F. Lentine, SNED - Zion M. Lesnet, Project Engineer I. Netzel,-Project Engineer S. Sanderson, Licensee Contractor B. Soares, ISI Enginee *J. Tiemann, Project Engineer D. Wozencraft, Project Engineer

• Indicates persons present at exit intervie . Licensee Actions on Previous Inspection Findings (92701, 93702)

(0 pen)_Open Item (295/86013-03; 304/86012-04) Reactor coolant system (RCS) vent. system not environmentally qualified per NUREG 0737. As

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indicated in-the attached Technical Review Report (TRR), this item is

under review by Region III for acceptabilit _(Closed). Unresolved Item (295/86013-04; 304/86012-05) Potentially inadequate documentation for pre-operational test' package for Unit 1 RCS vent system modification, M22-1-79-43. As indicated in the attached TRR,_the appropriate documentation was located and was found acceptabl '_(Closed) Unresolved Item (295/86013-05; 304/86012-06) Adequacy of i emergency operating procedures (EOPs) for RCS venting safety considera-tions. As indicated in the attached TRR, the licensee's response to this item is considered acceptabl .(Open) Open Item (295/85031-02; 304/85032-04) Conflict between Item E.1.c of the February 29, 1980 NRC Confirmatory Order and Technical l Specification (TS) Amendment No. 80/70, issued January 2,1983. As ;

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indicated in the attached TRR, the resident inspectors were to review recent plant operations for compliance with these requirements. Unit 2 l .

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y operations following the. August 1987 startup after a refueling outage

.were_ considered acceptable. The licensee intends to resolve this issue

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r by having Item E.1.c of the Confirmatory Order withdrawn based on TS ,

Amendment 80/70. This item is considered open pending the licensee's l review of the-potential Order /TS conflicts.-

l (Closed) Open Item (295/85005-06; 304/85005-06) Calibration test of the

.<3 containment high range radiation monitor in the 100 to 1000 R/hr rang As indicated in the attached TRR,-the results of this test were reviewed and found acceptabl 'No violations or deviations were identifie .;

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Inspection of Licensee Implementation of TMI Action Plan Items (25565) (

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At the request of NRC Region III, Parameter, Inc. performed an inspection

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li of the licensee's action for TMI Action Plan items. Details of that i inspection are contained in the attached TR ] Based on the attached TRR, the following TMI Action Plan items are considered closed: q I. Guidance for the Evaluation and Development of j l Procedures for Transients and Accidents '

I. Plant-Safety Parameter Display Console II. Reactor Coolant System Vents l

II.E. Auxiliary Feedwater (AFW) System Automatic Initiation and Flow Indication II.E. Containment Isolation Dependability II.F. Containment High Range Monitor II.F. Containment Hydrogen Monitor II.K. Automatic Power Operated Relief Valve (PORV) Isolation II.K. Automatic Trip of Reactor' Coolant Pumps During Loss of Coolant Accident II.K.3.25 Effect of Loss of AC Power on Pump Seals Item I.D.1, Detailed Control Room Design Review (DCRDR), will remain l open pending NRC review of the licensee's completed DCRDR l modifications.

l Regarding item II.E.1.2, AFW System Automatic Initiation and Flow Indication, the resident inspector reviewed recent plant operations for compliance with Zion Technical Specifications and Confirmatory .

Order Item E.1.c. No violations or deviations were noted. The I conflict between the TS and Order Item E.1.c remains open as described in paragraph 2 abov _ - _ - _ - _ _ . ._ _ _

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1. Regarding item II.F.1.6, Containment Hydrogen Monitor, the unsecured-hydrogen sample piping is considered an'0 pen Item pending NRC review of the licensee's corrective actions (295/87028-01; 304/87029-01). Regarding item II.K.3.5, Automatic' Trip of Reactor Coolant Pumps, Ethe inspector identified that the licensee had failed to submit an analysis'of the RCS wide range pressure. instrument uncertainties a .resulting from normal and adverse containment. This submittal should have been. provided by October 31,1985, as stated.in the

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licensee'.s: August 22, 1985 response to Generic Letter (GL) 8501 In~ addition, the. licensee's uncertainty analysis, approved by onsite review dated December 27, 1985, applied to Barton Model 763

. pressure transmitters, but the licensee is installing Rosemount Model 1149. pressure transmitters in both units. Thi s ' modi fication-is completed for Unit 2 and will be. performed during the nex '

refueling outage for Unit.1. _The licensee expects'to submit the required information by mid-November 1987. This is considered an

! Unresolved Item pending NRC review of the required uncertainty analysis for the replacement pressure transmitters, and NRC review of whether failure to submit the original analysis constitutes a deviation from an NRC commitment (295/87028-02; 304/87029-02).

No violations or deviations were identifie One Open Item and One Unresolved Item were identifie . . Inspection of Emergency Operation Procedures (EOPs) (25579)

As noted .in the attached TRR, while inspecting the licensee's response to TMI: item I.C.1, E0Ps were reviewed'for conformance with GL 82-33,

" Requirements for. Emergency Response Capability." This review also met most of the requirements of Temporary Instruction (TI) 2515/79,

" Inspection of Emergency Operating Procedures" Based on the satisfactory results of the review performed during this inspection, TI 2515/79 is considered close No' violations'or deviations were identifie . Open Items-Open Items are matters which have been discussed with the licensee, which wi'il be reviewed further by the inspector, and which involve some action on the part of the NRC or licensee or bot One Open Item disclosed during this inspection is discussed in paragraph . Unresolved Items

. Unresolved items are matters about which more information is required

. in order to ascertain whether they are acceptable items, items of noncompliance or deviations. One Unresolved Item disclosed during this inspection is discussed in paragraph _ _ _ . __

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7. . Exit Interview (30703)

' The inspector met with licensee representatives (denoted.in Paragraph 1)

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[throughout the-inspection period and at the conclusion of the inspection-on August 21,'1987,1to. summarize the scope and findings of the' inspection p . activities:as described in the attached TRR. The licensee acknowledged k , .the inspector comments. The inspector ~also' discussed the likel informational content _of the' inspection report with regard to documents or. processes reviewed:by the inspector during'the inspection. The

' licensee did not identify any such documents or processes as proprietar ~After reviewing the draft TRR, the resident inspector met with Mssrs. .

Schultz and A. Bless (see paragraph 1) on September 29, 1987 to summarize

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the NRC position regarding the status of Open Items, Unresolved Items and

'TMI Action Plan Items'as noted.in this report. .The licensee acknowledged the-inspector's comments. Proprietary material was adequately discussed during the August 21,.1987 exit meetin i I

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TECHNICAL REVIEW REPORT U.S.N.R.C. REGION III ZION NUCLEAR GENERATING STATION, UNITS 1 & 2 DOCKET NOS. 50-295 & 50-304 INSPECTION OF LICENSEE IMPLEMENTATION OF TMI ACTION PLAN ITEMS IN ACCORDANCE WITH NUREG 0737, " CLARIFICATION OF TMI ACTION PLAN REQUIREMENTS" TASK NRC-01E-157-01023 INSPECTION CONDUCTED MAY 13 - AUGUST 21, 1987 NRC TECHNICAL LIAISON PREPARED BY J. Hinds Parmeter, In M. Holzmer 13380 Watertown Plank R P. Eng Elm Grove, WI 53122 AUTHOR D. A. Beckman Prisuta Beckman Associates, In I

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TABLE OF CONTENTS PAGE INTRODUCTION......................................................... 1 I. Guidance for the Evalution and Development of Procedures for Transients and Accidents................. 2 I. Control Room Design Reviews................................ 4 I. Plant Safety Parameter Display Console..................... 4 II. Reactor Coolant System Vents.................... .......... 6 II.E.1.2 Auxiliary Feedwater System Automatic Initiation and Flow Indication............................. 9 II,E.4.2 Containment Isolation Dependability........................ 11 II.F.1.3 Containment High Range Monitor............................. 12 II.F.1.6 Containment Hydrogen Monitor............................... 13 II.K 3.1 Automatic Power Operated Relief Valve (PORV) Isolation..... 16 II.K.3.5 Automatic Trip of Reactor Coolant Pumps During Loss of Coolant Accident............................ 16 II.K.3.25 Effect of Loss of AC Power on Pump Seals................... 18 i

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TECHNICAL REVIEW REPORT '

INSPECTION OF LICENSEE IMPLEMENTATION OF 1 TMI ACTION PLANT ITEMS IN ACCORDANCE WITH {

NUREG 0737, " CLARIFICATION OF TMI ACTION PLAN REQUIREMENTS" l

ZION NUCLEAR GENERATING STATION INTRODUCTION Parameter, Inc., under the direction of NRC Region III, provided technical assistance to perform reviews and inspections for the below listed outstanding i Three Mile Island (TMI) Action Plan items as documented in this repor j The following items from NUREG 0737, " Clarification of TMI Action Pla ,

Requirements," were inspected in accordance with NRC Inspection Manual Temporary Instruction (TI) 2515/65 (Revision 1):

I. Guidance for the Evaluation and Development of Procedures for Transients and Accidents

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I. Control Room Design Review I. Plant Safety Parameter Display Console 11 Reacter Coolant System Vents

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II.E.1.2 Auxiliary Feedwater System Automatic l Initiation and Flow Indication li II.E.4.2 Containment Isolation Dependability i

II . F.1. 3 Containment High Range Monitor  ;

II.F.1.6 Containment Hydrogen Monitor II.K.3.1 Automatic Power Operated Relief Valve (PORV) Isolation II.K.3.5 Automatic Trip of Reactor Coolant Pumps During Loss of Coolant Accident II.K.3.25 Effect of Loss of AC Power on Pump Seals At the direction of NRC Region III, Item II.B.2.3, Environmental Qualification, and Item III.D.3.4, Control Room Habitability, were deleted i from the original scope of inspection. Item II.B.2.3 has been inspected '

separately by NRC Region III. The Control Room Heating, Ventilation, and Air Conditioning (HVAC) system (Item III.D.3.4) had also been inspected by NRC Region III, and was being evaluated for adequacy of the licensee's corrective actions and potential NRC enforcement action. Item II.E.1.2, '

Auxiliary Feedwater System Automatic Initiation and Flow Indication, was added to the scope of this inspection at the direction of NRC Region II I

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i l I As appropriate to the respecti.ve items, the following attributes were L inspected, on a sampling basis, per Sections 03.01 and 03.02 of TI 2515/65-q p, Program / Procedure Verification  ;

• i Programs'and procedures were developed and in conformance with licensee l commitments and NRC requirement '

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Programs 'and procedures were properly approved by the licensee before implementatio Up-to-date programs and procedures were available for us Personnel have been trained in the new programs'and procedures, and they understand and are using.the Equipment Installation / Modification and Operation 1

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Installat'on and modifications of equipment met licensee commitments and NRC requirement *

Equipment changes were properly approved and controlle !

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As-built drawings were changed to reflect equipment change Necessary procedure changes have been made, and personnel training has been accomplishe Pre-operational testing is complet *

Equipment is calibrate *

Equipment is operable, and operational procedures are being use Onsite inspection was conducted during multiple site visits by the assigned consultant during the period May 13 through August 21, 1987. Approximately 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> of onsite inspection were conducted. Persons contacted during the inspection are listed in Attachment References reviewed during the-inspection are listed in Attachment The results of inspection of each Item include a summary of the NRC requirements and/or licensee commitment and a discussion of pertinent discrepancies found during the inspection. Unless so noted, the other inspection attributes sampled above were found to'be satisfactor . Item I.C.1 - Upgrade of Emergency Operating Procedures (EOPs)-

In December, 1982, the NRC issued Supplement 1 (Generic Letter [GL]

82-33) to NUREG 0737, " Requirements for Emergency Response Capability,"

which included a requirement to upgrade E0Ps. GL 82-33 required implementation of human factored, function oriented E0Ps, including formalized programs for verification of technical accuracy and valida-tion in the form of a Procedure Generation Package (PGP).

The licensee responded to GL 82-33.by letters dated July 20, July 28 and August 14, 1983, which provided its plans and commitments for preparation of the PGP. The licensee participated in the Westinghouse Owner's Group (WOG) for the development of generic Emergency Response Guidelines (ERGS)

subsequently used as the basis for plant specific procedure ;

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On' June-12, 1984, the NRC issued a Confirmatory Order to the licensee confirming the milestone schedules for completion of E0P (and other GL

'82-33) activities. -The PGP was subsequently submitted on December 28, 1984, and the actual procedures were implemented in late 198 The PGP consisted of four procedures (ZAP 5-51-4A, 48', 5, and SA; see Attachment 2) and.provided requirements for procedure content,' format, verification and validatio The inspector reviewed a sample of.the above procedures with respect to the requirements of NUREG 0737 and GL 82-33 and verified conformanc A general review of the E-0, E-1, and E-3 series E0Ps further confirmed that the plant specific procedures generally conformed to the licensee's PGP procedures. The inspector also reviewed conformance of the plant

, specific procedures with the WOG ERGS, including adequate disposition of deviations from-the ERGS identified in the licensee's submittal to NR During this general review, the inspector also used the guidance of NRC Information Notice (IN) 86-64 (including Supplement 1), " Deficiencies in Upgrade Programs for Plant E0Ps." This review included the licensee's .

evaluation of IN 86-64, documented in Action Item Report 295-103-86-06 j E0P Series E-3, " Steam Generator Tube Rupture," including Emergency Subprocedures (ESs) 3.1-3.3, was reviewed in more detail, as well as Emergency Contingency Action Procedures (ECAs) 0.0S 3, Functional Restoration Procedures (FRs) 5.1-5.2, C.1-C.3, anc h 1-H.5, and Status Trees F 0.1-0.6. These procedures were reviewed, on a sampling basis, with respect to the WOG ERGS, plant specific design features, and the requirements of the PGP procedures. Portions of the procedures were compared with plant equipment, instrumentation and controls during plant tours. All inspector questions were adequately resolved by the cognizant Operating Enginee See Item II.K.35 elsewhere in this report for a discussion of discrepancies in the Reactor Coolant Pump (RCP) trip criteria used in the E0P Records of the procedure validation and verification program were also reviewed (see Attachment 2) for the E-3 series procedures, confirming that the. licensee had implemented this portion of the PGP progra l The inspector confirmed that procedure discrepancies and improvements identified during this process were properly evaluated and incorporated into the E0P i Training records (Attachment 2) were reviewed on a sampling basis to ensure that pre- and post-implementation training had been provided to :

the operators on the principles of the procedures, on specific procedure i performance, and on procedure implementation on the plant simulato l Specific training was verified for six licensed operators for RCP trip l criteria, E-3 Series classroom and simulator training, Safety Parameter 4 Display System (SPDS) use training, and containment high range radiation monitor classroom training. The inspector also confirmed that questions, l procedure / plant comments, and other feedback sources identified during training were appropriately reviewed and dispositioned by the plant operating staff.

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c This item is recommended _for closure by the Resident Inspector . Item I.D.1 - Detailed Control Room Design Review (DCRDR)

NUREG 0737, Item I.D.1and GL 83-22 provided requirements for the

' performance of a human factors based DCRD Prior activity on this item included implementation of the DCRDR by the licensee, an in process audit by the NRC Office of Nuclear Reactor Regulation (NRR) and its consultants,- resolution of NRR review and audit comments by the licensee, and issuance of an NRC Safety Evaluation Report (SER) on March 10, 198 Attachment 2 lists pertinent reference At.the time of this inspection, the licensee had established and the NRC staff had accepted (with exceptions) a schedule for implementation of modifications resulting from human engineering discrepancies resulting from the program. With the exception of a small number of plant modifications (control room central work station, for example), essenti-

' ally nene of the modifications have been implemented; they are scheduled for implementation during the next two refueling outages (1988-1990).

As a result, no post-implementation review of licensee DCRDR modifications was conducted during this inspectio he inspector further noted that the current outage schedule may have slipped (due to protraction of the recent outage) from that originally-provided by the licensee's May 1, 1986 schedule proposal and accepted by the NRC's March 10, 1987 SER. The licensee's Regulatory Assurance representative was apprised by the inspector of the need to ensure that the previously docketed commitments are either met or revised dates submitted for NRC approva This Item is open pending NRC review of the licensee's completed DCRDR modification . Item I.D.2 - Plant Safety Parameter Display System (SPDS)

NUREG 0737 and GL 83-22 provided requirements for provisions of the SPDS to provide a concise display of critical plant variables to control room operators during abnormal and emergency condition The requirements included specific design criteria and features, minimum parameter selection criteria, human factors design considerations, and a system verification and validation program. The licensee was required to submit a Safety Analysis describing compliance with the above for NRC staff

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revie The Safety Analysis was submitted via Commonwealth Edison Company letters of December 29, 1983 and February 10, 1984. The NRC staff review was documented in Safety Evaluation Reports dated June 22, 1984 and June 26, 1 1986 (with comments and exceptions noted). Implementation of the i licensee's SPDS program was the subject of a Confirmatory Order issued by ]

the NRC on June 12, 1984, confirming that the required submittals to the

, NRC had been made, that the system was in operation, and that operators

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,1 The final report of the licensee's SPDS Verification and Validation (V&V)

Program was issued on July 8, 1987 and reviewed by the inspector; the )

results of this review are provided below. The inspector also observed <

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the display and control installations, confirmed the provisions for 1 operator training, and reviewed integration of the information into 'l control room normal and emergency operation With regard to integration into plant operations, the inspector noted that no discrete procedure for use.of the system or its information was available. Although operation of the system is included in the Process Computer System Description and training material, and the operators appear sufficiently familiar with the system and its use, the use of the system is not otherwise integrated into plant procedure # The licensee noted that this finding had been identified by the V&V Program Human Factors Surveys and that a User's Guide was under preparation and scheduled for issuance during 1987. The inspector reviewed the expected contents of the guide and found it acceptabl The inspector was also advised that the plant simulator does not currently include replication of the SPDS, but that a simulator modifi-cation was under design. Discussion of operator training verification is provided under Item I.C.1 of this repor The inspector reviewed portions of the V&V Plan implementation and, where feasible, compared its results with the system and support documentatio 'Several discrepancies and inspector questions resulting from this review were adequately resolved by the licensee as follows:

The V&V Plan included extensive checklists for various discipline design reviews (e.g. operations, computer, maintenance, human factors,etc.). Many of the checklists included items which appeared to have no specific responses documented by the reviewe The licensee stated that the checklists, prepared by 'a consultant, frequently exceeded the licensee's intended scope of review or included items inappropriate to the specific discipline reviews (such items also appeared on the appropriate discipline checklists),

thereby warranting the reviewer's determination of inapplicabilit For a sample selected by the inspector, the licensee demonstrated that the key elements of each checklist were adequately covered by the overall program or were indeed inapplicabl The inspector further noted that all review checklist comments were not indiv? dually dispositioned. The licensee provided the inspector with a series of reports documenting the review and disposition of the comments by a steering committee, demonstrating that the

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comments had received appropriate levels of attentio '-

The design documentation included in the V&V Report (Volume II, Item 2.10) included informal comments from the station computer staff  ;

which indicated that the SPDS method for selecting core exit

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temperature thermocouple used in the subcooling margin calculations and the method of calculation applied by the SPDS computer program were inadequate. No resolution of the comments was apparen ______- _ _____

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At the inspector's request, the licensee provided a response to the l comments, dated March 15, 1983 which indicated that the comments had been addressed and appropriate modifications made to the SPDS ]

computer program. Excerpts from the revised program were included 1 as an attachment to the mem :

This item is recommended for closure by the Resident Inspector . Item II.B.1 - Reactor Coolant System (RCS) Vents NUREG 0737 requires installation of RCS and reactor vessel (RV) head I high point vents remotely operated from the control room. A description of the design and procedures and supporting analysis for use of the vents was required to be submitted to the NRC staf Item II.B.1, Clarification A.(10) requires that the vent system be environmentally qualified in accordance with IEEE 344-1975, relevant NRC references, and the Conmission Order and Memorandum (CLI-80-21, May 23, 1980).

'The licenses submitted information relating to RCS vent design to NRC via letters dated July 1,1981 and May 17, 1982. The NRC SER on this matter was issued cn September 9, 1983, and accompanying Technical Specification Amendments were issued on February 5, 198 This item has been previously inspected by NRC Region III as documented in Inspection Report 295/86013; 304/86012. This inspection addressed the acceptability of the installation, testing, maintenance, and procedures associated with the vents and identified several Unresolved and Open Items which were reinspected as described below, Open Items 295/86013-03; 304/86012-04 - RCS Vent System Not Environmentally Qualified per NUREG 0737 The prior inspection noted that the RV Head Vents were not included in the licensee's environmental qualification (EQ) Program. During this inspection, the inspector again reviewed the licensee's docketed EQ Program and again found that the head vents were not include The EQ Progrem, Appendix G, provides the EQ status and data for NUREG 0737 required modifications (e.g., containment monitoring instrumentation); the RCS Vent system was not included. Similarly, the valves and accessories were not listed as approved EQ items, and relevant suppliers were not identified as approved EQ source The licensee orally confirmed that the system is not environmentally qualified and, further, is not categorized as safety relate A conference call between the inspector, site Regulatory Assurance ,

personnel, and Station Nuclear Engineering Department (SNED) j management provided the following licensee position on the matter: )

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(1) Although NUREG 0737, Item II.B.1, states that the system be )

environmentally qualified per CLI 80-20, Commonwealth Edison j stipulates that the NUREG 0737 requirements (published I

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November 1980) predate and are superseded by 10 CFR 50.49,

' " Environmental Qualification of Electric Equipment Important

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to Safety for Nuclear Power Plants."

(2)- The Commonwealth Edison EQ Program, issued April 10, 1984, is in-compliance with 10 CFR 50.49, as demonstrated'by its

acceptance by the NRC.and by the results of special-NRC Inspection 295/8506; 304/8506, which found'that the licensee's methodology and implementation of the program were sound, including the screening criteria applied to candidate equipmen '(3) The September 9, 1983 NRC SER for. Item II.B.1, RCS Vents, further stipulated that the NRC's evaluation of the licensee's actions did not' include evaluation or accepte.nce of environ-mental qualification (and other) considerations. The SER stated that EQ was subject to a post-implementation NRC audit-in conjunction with other on going actions or programs and would be reviewed separatel Based on this.information and the NRC inspection results discussed in (2) above, the licensee considers the EQ program to be acceptable to the NR (4) The RCS vent' system was not included in the prcgram because it did not meet the screening criteria for inclusion, and no

.further action is deemed necessary by the license It should be further noted that the licensee was unaole to demonstrate by documentation or other means that the subject equipment had been subject to the screening process or that the results of that screening' process warranted exclusion from the EQ progra The vents would be operated after accidents per E0P FR I.3,

" Response to Voids in the Reactor Vessel," Revision 0, during scenarios where adverse environmental conditions will likely exist The licensee maintained that the need to operate the vents would ,

occur relatively early during the accident scenarios, thereby reducing any adverse environmental affects. The licensee further noted that the above procedure is a contingency action which is not i considered a true " design basis" scenari However .the procedure does not include guidance for operators which would ensure that the valves are not operated in environments for which they are not qualified and which thereby would prevent aother f leakage paths should the valves fail ope At.the close of this inspection, this matter was under review by NRC Region III management.. This item will remain open pending  ;

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i b. Unresolved Item 295/86013-04;'304/86012-05 - Potentially Inadequate Documentation for Pre-operational Test Package for Unit 1 RCS Vent System Modification (M22-1-79-43) l

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not have approval signatures prior.to use,-and did not identify

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the date'of test performance or the person performing the test, During this inspection, a review of the archival microfilm copies

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of the subject. test packages for Unit 1 and Unit 2' confirmed that the above signatures and information were included, q The inspector noted that the package records were microfilmed in disjointed order but appeared complete with respect to the test ;

records. This item has been recommended for closure to the NRC '

Resident Inspector c. Open Item 295/86013-05; 304/86012-06 - Adequacy of Licensee E0Ps for RCS Venting Safety Considerations ]

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During the prior inspection, the E0Ps did not include instructions i or guidance for ver; ting that addressed safety considerations such f as pressurizer relief tank integrity, initial containment hydrogen i concentration / buildup, operation of the reactor containment fan '

coolers (RCFCs).during venting, and operation-of hydrogen _ control equipment to remove hydrogen from containment. These considerations were included in a licensee response to NRC questions on Item II. as factors to be addressed in the E0P As part of the review of Item I.C.1, Emergency Operating Procedures, the inspector reviewed the E0Ps for loss of heat sink, inadequate core cooling,~and response to voids in the reactor vessel. These i procedures are representative of conditions which would require venting the pressurizer or reactor vessel to establish flow through i or remove voids from the cor / 1 The inspector noted that E0P FR I.3, " Response to Voids in the Reactor Vessel," Revision 0, includes assessment of containment j hydrogen concentration, closure of purge and exhaust valves, {

operation of RCFCs, hydrogen build-up vs. venting time, and 1 termination criteria. Other E0Ps, such as E0P FR H.1, " Response i to Loss of Secondary Heat Sink," Revision 0, require establishing i an RCS feed and bleed path by opening the pressurizer power operated relief valves and venting coolant back to the containment via the pressurizer relief tank. This and the other procedures (e.g. F-0.2, Core Cooling; F-0.3, Heat Sink Status Trees) reviewed did not include the same provisions as for venting via the RV head vent The inspector discussed the above with the cognizant Operating i Engineer and reviewed the WOG ERG Background Document, finding that:

(1) Voids in the reactor vessel are categorized as a " Yellow" l condition Critical Safety Function (CSF). j

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(2) Loss of heat sink and the other similar scenarios requirin L pressurizer. venting are. categorized as " Red" condition CSF .(3) .The'WOG ERGS and the WOG Executive Background Document distinguish between the Red and Yellow CSF conditions.in.that the Red condition CSF requiresLmaximum. response, and its greater safety priority overrides considerations of; containmen >

conditions, the Yellow condition CSF permits consideration of prerequisite or lower threshold caution '{

The licensee's position on this matter is. therefore considered j

! acceptable..and appears consistent with the NRC approved WOG ERG l These. items were recommended for closure to the NRC Resident i

~ Inspector Another Unresolved. Item (295/86013-06; 304/86012-07), involving reliability of RV head vent position indication, was not inspected during this review; this item is being followed separately by the

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. Resident Inspectors in conjunction with related Inservice Testing (ASME XI) is' sue This TMI item is recommended for closure by the Resident Inspector . Item II.E.1.2 - Auxiliary Feedwater (AFW) System Automatic Initiation and Flow Indication NUREG 0737 required that AFW Systems include safety grade automatic ~

initiation and flow indication features. The initiation features are required to include:

Single failure design considerations

• .

Testability.of initiating circuitry

• Emergency power source

Sequenced loading on emergency buses

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Retained manual' initiation capability-Manual capability in the event of failure of automatic circuits The flow indication is required to be safety grade and powered from j emergency buse ~'

Licensee submittals of December 14, 1979, December 15, 1980, December 31, 1980, July 1,1981, July 10, .1981, and April 15, 1982, provided the Commonwealth Edison response to this item, as required by NUREGs 0737 and 0578 and an NRC letter of September 18,1979 (Eisenhut to Reed, NRC i

Requirements for Auxiliary Feedwater Systems At Zion Station). The NRC SER was issued in conjunction with Technical Specification Amendments i 80/70 (Units 1/2) on January 21, 198 The licensee's correspondence stipulated that the automatic initiation and indication functions met. the NRC requirements as installed at the time of the initial submittals. The letters provided descriptions of the features which satisfied the criteri l

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? This and related NUREG10737 items were previously inspected during NRC Inspections 295/81009; 304/81009 and 295/85031; 304/85032. An-Open-Itemrresulting from the latter inspection is further disc'ussed belo " '

The7 inspector' reviewed as-built drawings, instrument calibration

. procedures and= data, plant operating and emergency procedures as listed

.in* Attachment.2,'and inspected in situ equipment during, plant tours.to confirm the licensee's functional descriptions. No discrepancies with

' respect to the. licensee's'submittals or_NRC SER were identifie ' ~

.0 pen Item 295/85031-02; 304/85032-04 involved a conflict between Item ir E.1.c of the February 29, 1980 NRC Confirmatory Order and Technical- ,

-Specification'n Amendment No. 80/70, issued January 21, 1983. ,

Order Item E.1.c required.the licensee to impose.an'adm_inistrative order requiring " expeditious shutdown whenever an independent train of the AFW i Lsystem and'any.one.of,the following are inoperable: all backup sources of offsite power, one of the two diesel generators supplying power to the other independent train, or either of the other trains of the AFW system."

A~ conflict between = the' above r.nd TS 3.7.1.2, Auxiliary Feedwater System

.(Amendment 80/70) was identified to the NRC by a December 6,1983 -

Commonwealth' Edison letter, Barnes to Denton. This letter stated that certain provisions of the TS were in conflict with the Order but did not specifically identify the provisions or conflict ( f Discussions with the Product' ion Superintendent, Regulatory Assurance ?):I

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Supervisor, and a Shift Supervisor confirmed that the operations Standing ,

Order originally issued for Order Item E.1.c had been rescinded by plant-management at the time of the TS issuance on the basis that the amended TS included requirements equivalent to the provisions of the Order. The site staff.was unaware of and unable to determine the specifics which resulted.in the December 6, 1983 letter to NRC. The staff's evaluation t

of the matter was continuing at the close of this inspectio Upon consultation, the NRC Licensing Project Manager (LPM) advised that:

only two -items of the Confirmatory Order (A.1 and A.2) had been_ rescinded by the NRC, that the order formally modified the license conditions and,

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therefore, that it takes precedence over the TS and other license 3 conditions. The LPM advised that provision E.1.c of the order remained in full effec This inspector,'in conjunction with the Resident Inspectors, reviewed the current Zion TS Sections

3.7.1.2, AFW System, 3.15, Auxiliary Power Systems, and 3.0.3 and 3.0.5, General Limiting Conditions for Operatio These TSs provide actions which appear to be equivalent to those stipulated by the order, with action time limits specified for all except loss of all offsite power, except as noted belo With all three AFW pumps inoperable, the TS does not require $mmediate  !

shutdown, based on not perturbing the plant and thereby inducing conditions which would result in an AFW challenge.' The licensee considered that this TS is in conflict with the aspect of the Confirmatory Order which requires expeditious shutdown with one train and either of the other trains of AFW inoperabl l l

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'LThe T5s appear to adequately address the following conditions. With

yl y' i more than one train of AFW' inoperable, restoration must occur within ( 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or the unit must be placed in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. With one train of AFW inoperable and a diesel generator inoperable for the redundant train, TSs. 3.0,5 and 3.0.3 apply, requiring '

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'7 testorstion within one hour or placing the unit in hot shutdown within

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the next.four, hours. With no offsite power supply lines operable, the unit must be placed in hot shutdown with no time limit specified 1 (discussions with the plant staff indicated that the time limits of '

TS 3.0.3 would be applied). (

AlthoughthelicenseeappearstobeintechnicalvN1ationofItemE. of the Confirmatory Order due to the cancellation of.the station {'

administrative order, operation in compliance with the above TS would appear to meet the requirements of E. The Resident Inspector was separately reviewing the licensee's recent operating history with regard to simultaneous diesel generator - AFW <

inoperability to confirm compliance with both the above TS provisions and the Confirmatory Orde ;

This item is recommended to remain open pending completion of the j licensee's review of potential TS-Order conflicts and the Resident !

$, Inspector's review of recent plant operation I Item II.i!.4.2 - Containment Isolation Dependability  !

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NUREG 0737 included seven principal positions involving:

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1 Diversity of sensed parameters for initiation of containment isolatio ] Identification and justification for selection of essential and nonessential systems isolate . Isolation of all nonessential systems by the containment isolation signa . Valves will not automatically reopen upon isolation signal rese l Minimum practical containment pressure setpoint for isolation of l nonessential system _

l Containment purge valves must meet current NRC requirements or be sealed close . Containment purge and vent isolation valves must close on a high radiation signa The licensee's responses to this item were provided by letters of December 14, 1979, December 15, 1980, December 31,1980, July 1,1981, I and April 15, 1982. NRC SERs were issued on February 28, 1980, 1 September 9, 1931, September 30, 1982, and April 3, 1984. See Attachment i '

Licensee actions with regard to Items 1, 5, and 6 were inspected during NRC Inspections 295/81001; 304/81001 and 295/,81009; 304/81009.

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v f <. was confirmed, on as a pl Wa b.' sis,

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g to be in accordance with the d ,various licensee suuntQ1s hia review of af-built drawings, instrument

' , e ' calibration data,,opI, rating ind emergencv procedures and modification

• packages. Also, implementation of isolation actuation and containment

,'f 1 isolation and integrity Technical Specifications was reviewed on a l

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1. sampling basi T The inspector veriN ed that the. Containment Isolation Phase A & B schemes

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, and setpoints in effect?were consistent with the submittals and SER Modification M22-1/2-79-13, which corrected; isolation reset deficiencies, was reviewed and found acceptable with respect to Items 3 and '

' Modification of and operating' procedures for the containment purge and i

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exhaust valves were. reviewed, pnfirming them go be consistent with the l

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requirements and NRC guidance. Y No discrepancies were identifi '

by the Resident Inspector +Y i f Qitem .s recommended for closure+ ,;

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cs ' Jjem II F.1.3 z - Containment High_Jjakge_ Monitor i g3, 3 s

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'lVREW6f 37greq11' red installation of $ntainment radiation monitors wih a maximum'rpnge of 10rdo/hr and provided specific design padameters

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A . and gi.idancV in T.ible II.F. ,

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• The 'sicensee committed to install suchbacnitors and described its  !

pvN 31, 1980, December 15, 1980, July,Pam in'latters 1, 1981, do the December 6, NRC 1983, dated and March Daember 21, 1985. An NRC letter of January 8, 1982 confirm,ed comp 1'etion of an NRR review of w e licensee's

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lInensee actions with v4 gard to this item were previousiv inspected in i

\Nkt' Inspections 295/82712; 304/82016 and 295/85005; 304/8500 An open ihm reMilng f rom the latter inspection is further discussed belo . ,# r The inspector reviewed the modification package, vendor test reports, invironwital cualf.fication reports, operating and emergency procedures,

' knd drawings, and cd. firmed that the equipment met the design requirements ofNUREGg737and,thelicensee'ssubmittals(seeAttachment2).

s During thid review, the inspector noted that, in a December 15, 1980 letter to the: NRC, lhe licensee notified the NRC that the lower limit of linearity o/ tN/ detecthrs iould be 85 Kev vice the 60 Kev of NUREG 0737, v Table II.F.M. Od re@w of venc'or energy response data (Report

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E-255-958, s e Attachme % 2),N the W yecter found that the sensitivity curves for the detector'had pointt lof deflection 'n the range of interest. The ' report wc h f ient rebarding evaluation af the curves (it

, appeared to be missing the.page(s) which discusse1 them).

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At the inspector's request, the licensee ct.it' acted the vendor and obtained the missing page, which included a statistical analysis of the data ' points d?monstrating linearity fron @0 Kev to 3 Kev with a 40%

% median or +/-POVlinearity.

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H _ Table II.F.1-3. requires that,J prior to initial use, calibration of each unit'must' include calibration with a radiation source at at least one ,

4 , point per decade in the range;of.1 R/hr and 1000 R/hr. In a letter dated

• March 21,'1985,- the licensee. notified the NRC that the monitors had not

'been so calibrated in the'100 to 1000 R/hr range (but had been vendor '!

calibrated-in the 1 to 100 R/hr decades). The letter further provided

% commitments for completion of this testin f k view of licensee completion'of the above' test identified Open Item J

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L 495/85005-06; 304/85005-06 during a previous. inspection. The inspector reviewed. Test Procedures TSSP-95/96-85 performed on May 22, 1985 for

.Un.it 2 and December 5, 1985 for Unit'l and found that the data met the-above commitments. Closure of this item is recommended to the Resident R6 wy ; ye'

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Inspectors,

' Item II.F.1.6 - Containment Hydrogen Monitor NUREG 0737. required' the capability for continuuus control room indication of/ hydrogen concentration (to 10%) in the containment atmospher *

The Commonwealth Edison letters of. December 15, 1980, February 8, 1983, and May'13, 1983 provided the licensee's commitment to install continuous monitors. The NRC SER'was issued on September 9,1983, documenting

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acceptability of the licensee's design as described by the submittal The' inspector reviewed the pertinent references in Attachment 2, y . inspected the in situ plant equipment, and confirmed, on a sampling R

( basis, that the installation and operation of the system were consistent with the licensee's submittals and the NRC SER.

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During review of ST-23-81, Hydrogen Analyzer Test, performed December 1981 - July 1982, the inspector identified several test discrepancies

'for which the resolutions appeared ambiguous:

In Step 7.9, testing of the cell failure alarm, the procedure was '

annotated to indicate that a calibration resistor of a value sfferent than that specified was required to obtain satisfactory test results. The cognizant Technical Staff engineer and Instrument Maintenance personnel were initially unable to establish the reason

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- for the above annotation and.the basis for its acceptabilit Difficulty was encountered in obtaining adequate schematic diagrams i

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s to evaluate the circuitry affected, and the actual location of the

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resistor installation could not initially be determine )

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The licensee later established that the resistor was installed as a temporary test aid to artificially load the circuit to test an  :

alarm. The inspector was provided a Station Procedure Change l Request (approved November 9, 1984) for the permanent calibration procedures which replaced the use of the load resistor with a DC

, voltage source.

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In Step 7.15, containment-to-analyzer flow path verification, the l procedure was annotated to indicate that the Unit i valve controls L 'were improperly wired,. apparently the result of an AE drawing erro The notes indicated that the valves were field rewired and that the )

test was accomplished satisfactorily. No documentation was provided in the modification package to demonstrate that the changes received engineering approval and that the drawings were corrected. The inspector verified that the rewiring and drawing corrections were properly processed via Work Request 18826, dated October 21, 1982, and that the drawing revisions were mad During an inspection tour on July 8-9, 1987, piping attached to the hydrogen monitor suction and return lines (1/2" socket welded, stainless

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steel) was found unsecured. The piping branches from tees in the monitor lines (about ten feet above floor level) and runs horizontally (about i ten feet) then downward (about six feet) to a manual sample valve ;

manifold (five to six feet long). The sample manifold is also connected to the service air system and a containment post-accident radiation monitor sample pump. Unit 1 and 2 configurations are nearly identical and both are affecte The horizontal portion of the piping run is supported by U-bolt clamp l The vertical (downward) run.and the sample valve manifold is intended to be supported by removable sample bombs flanged to the manifold and to a floor mounted support " table."

The sample bombs are missing, leaving the entire assembly cantilevered from the socket weld tees in the main piping run, the sample pump, and the U-bolt hangers. The valve manifold can be manually moved (swung) a nominal four to six inches with minimal hand force, placing what appears to be significant stress on the piping and weld Commonwealth Edison advised that the branch piping had been " retired" and that the sample bombs were removed for disposal or other us l Another line from the manifold runs to a flanged connection on a containment atmosphere post-accident radiation monitor located adjacent to the sample table. The piping-to-flange weld joint is also in torsion when the manifold is move A portion of the manifold is connected to a plant air header via double !

manual isolation valves; the piping interconnections partially support i the manifold. The air piping is supported only in the vertical direction by a turnbuckle hange Although various segments of the manifold and connected piping are isolable via manual valves, the configuration is sufficiently complex that multiple flow paths can exist from nearly any potential break ,

locatio TS 3.8.8.B requires that two hydrogen monitors (per unit) be operabl With both monitors inoperable, at least one monitor must be restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or the plant must be placed in Mode 3 within the next six hour __

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On July 9 and 10, 1987, the licensee . was requested to provide the seismic

' design requirements _for the. piping, to provide its position on the-

.. ope rability of-the piping during seismic events (with respect to

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Technical Specifications), and to provide its intentions for correction of th'e condition. The onsite information presented on the piping diagrams and the. Architect Engineer (AE) Piping Design Table, Section E.1, was ambiguous such that the plant staff was unable to establish the actual seismic classification. This request was referred to SNED. On i July'10, 1987, the licensee advised the Senior Resident Inspector that !

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it believed that. the piping design does not require seismic qualification

- and that piping operability is therefore not compromised. 'The licensee stated that this preliminary information would be confirmed via a i'

documentation revie No further information was provided by the licensee until August 17, 1987, when the. licensee was again queried about the status of this j matter. The licensee provided.two letters from its AE (December 3, 1980 d and October 21, 1981) indicating that the piping was designed as at least Seismic' Class-II. The licensee further advised that no additional action had been taken to reinstall the sample bombs or otherwise support the branch, pipin Seismic Class II is. defined by the Zion FSAR, Section 1.10.1.1 as designed to.DesignLBasis Earthquake standards via static (vice dynamic)

t analysis methods and applies to "... components not Class I which function in direct support of reactor operation."

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When the NRC questions of July 8-10, 1987 were reiterated to the licensee on' August 17, 1987, the inspector was orally advised of the following

'SNED position: Although the AE letters above indicate the piping is designed to Seismic Class II,' Commonwealth Edison does not consider that designation to be a " regulatory design requirement or commitment." Based on 1 above, Commonwealth Edison considers conformance with Seismic Class II design standards to be at Commonwealth Edison's option; i.e., Commonwealth Edison is not obligated to maintain a i Seismic Class II system configuration and the absence of same does not affect Technical Specification operabilit I The piping configuration is being reviewed by the AE (begun in July 1987), but this review and any attendant analysis have not yet been completed or reported to the station. A permanent modification for eventual removal of the piping (or equivalent) has been initiated;

.the implementation schedule is under Commonwealth Edison review and may be accelerated to address this issu . Reinstallation of the sample bombs is not feasible; apparently none are availabl . Commonwealth Edison believes that a temporary modification is not necessary'and that, in any case, it would be contrary to its current modification polic _ _ - - _ _ - _

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L .It'should 'be noted that USNRC. Reguloatory Guide 1.29 (not a Commonwealth L Edison commitment-or requirement) requires that containment hydrogen i- control systems 'and systems' used .to monitor or actuate same should be considered Seismic Category The licensee was advised that'the above issue.is not one.of regulatory requirement; applicability,- but rather the functional. operability of the f piping:under.a' seismic event such that the principal containment hydrogen monitoring flow path is not degraded. Additionally, the effectiveness of

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the configuration' management programs which require the application of the:above Commonwealth Edison position is cause for concer At the' exit meeting for this inspection on August 21, 1987, the licensee provided the inspector a memorandum from the' Commonwealth Edison Nuclear

! Licensing Administrator to'the Plant Manager (dated August 21,1987)

. generally confirming the above. licensee regulatory pos'itions. Additionally, 1 it'noted that Commonwealth Edison will be required by previous NRC initiatives'to impose the' seismic criteria of Regulatory Guide 1.97 ,

pending' resolution of.open issues with NR At the close of this inspection, the matter of the unsecured hydrogen i monitor. piping had been referred.to NRC Region III management for evaluatio This TMI item is recommended for closure by the Resident Inspector ! Item II.K.3.1 - Automatic PORV Isolation NUREG 0737 re' quired'all PWR licensees to provide a system that auto-imaticallyJcloses the PORV block valve to protect against small break loss of coolant' accidents from PORV closure failure The. licensee had adopted the results of a Westinghouse Owners Group (WOG)

effort taken in response to the above item and, in letters of April 1, 1981 and July 1, 1981, advised the NRC staff that no mo(ifications to incorporate the subject features were necessary or propose Via an SER, dated September 12, 1983, the NRC staff accepted the conclusions of the WOG generic report and found that no modifications were required based on the intent of Item II.K.3.2 being met with the existing PORV, safety valve, and reactor high pressure trip setpoint l The inspector confirmed that the above plant parameters were as stated 1 in the submittals and SER by review of current Technical Specifications and operating and calibration procedures as listed in Attachment 2. No discrepancies were identified. This item is recommended for closure by the Resident Inspector . Item II.K.3.5 - Automatic Trip of Reactor Coolant Pumps (RCPs)

l NUREG 0737 required the licensees to install modifications to =

automatically trip RCPs to lessen the severity of small break loss of coolant accidents. This modification was to be accomplished while 16 i

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L . industry and NRClresearch and evaluation programs continue In February 1983,' following analysis of Loss of- Fluid Test Facility test-

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, data, the:NRC issued Generic Letters'(GL)83-10c and d requiring plant 3

. specific. analysis and development of.RCP trip setpoints sensitive to the -broad range of accident spectra.

i 'The~ licensee participated in Westinghouse Owners Group (WOG)' development of.a generic. response'to GL 83-10c and d, which concluded-that'an

automatic trip of the RCPs was not required based upon development of plant specifice alternative manual trip criteria based on RCS pressur GL 85-12 was issued June 28, 1987, requiring the licensees to select, implement, and provide NRC-with the bases of' acceptability for the plant

. specific criteria. . Wide. range RCS pressure instruments inside contain-ment are used to determine the plant pressure for manually tripping the RCP The' licensee' responded to GL 85-12 in letters dated August 22, 1985, and

' September 18, 1985,- responding to all but one NRC information requiremen GL-85-12,: Item A'2 required the licensee to submit analysis of the wide range RCS pressure instrument uncertainties resulting from normal and adverse containment environment The inspector'. reviewed the licensee's.submittals, the Emergency Operating Procedures (see Item I.C.1 earlier in this report) and operator training and confirmed that the licensee had incorporated the criteria into'its

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program The licensee's. letter of August 22, 1985 stated that this information would be'provided separately by October 31, 1985. The licensee's subsequent letter of. September 18, 1985 included a Section A.3 which

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discussed RCP trip criteria, but the specific information required by GL-85-12, Section A.2 was not included. When advised of the above by the inspector during the week of May'13, 1987, the licensee confirmed

.that the information-had apparently been overlooked in the latter submittal, but would be submitted expeditiousl The above, failure to submit information required by GL 85-12 in accordance with the August 22, 1985 commitment to NRC appears to represent a Deviation. This matter was referred to NRC Region III management-for dispositio The inspector was subsequently provided with Onsite Review Committee Package OSR/08~9/85, approved December 27, 1985, which' included the uncertainty analysis and relevant data that had been incorporated irJo 4

.the Emergency Operating Procedures. This data, although not transmitted !

to the NRC, appeared to meet the information requirements of GL 85-1 The inspector reviewed the calculations and enalysis with respect to the i vendor instrument data, Emergency Operating Procedures, and the WOG analysis and Emergency Response Guidelines, finding them consistent with NRC and WOG guidance,

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During the week of August 17, 1987, the inspector was further advised

.that the above submittal had not yet been made. During the last outage, the previously. analyzed Unit 2 wide range RCS pressure transmitters (Rosemount: Model 1149). had been replaced-(Modification Package 86-38)

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.with (Barton 763) units of_different manufacture, invalidating the prior

. analysis. The ~ formerly' installed instruments had -a history of- excessive drift. . The Unit 1 transmitters are' also scheduled 'for replacement during the 1988 outag An-uncertainty analysis for the new instruments ha'd been received from the' plant AE;and was under licensee review. That analysis indicated a small but material change in the adverse environment uncertainties which would result'.in a small downward change in the RCP trip pressure t

e cri teri a .' ' Revisions to the E0Ps were also in progress but not yet complet The inspector reviewed the changes in trip pressure with the cogn'izant Operating. Engineer and determined that the values currently in the E0Ps (from the prior transmitters) were generally conservative with respect to the new transmitter criteria and appeared acceptable until the E0P revisions were routinely processe At the close of the inspection, the above E0P revisions were still in

. process'and the required submittal to the NRC was still pendin . - ' Item II.K.3.25 - Effect of Loss of AC Power on RCP Seals This' item required that the licensees determine the consequences of a

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loss of cooling water to RCp seals to demonstrate the adequacy of the seal design to sustain a complete -loss of AC power for at least two hour .In letters dated December 15, 1980 and January 8,1982, the licensee stated that, in the event of a loss of offsite power, component cooling

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water to the thermal barrier heat exchanger and chemical and volume control seal injection flow are both automatically restored within 30 seconds upon automatic starting and loading of the emergency diesel generator The letters also note that upon loss of offsite power, the RCP' motor itself is.de-energize Further, the RCPs are designed to function for up to 30 minutes without cooling and seal injection and require only seal injection or thermal barrier cooling to prevent seal failure for up to two hour The above position was accepted by an NRC SER issued May 26, 198 The inspector reviewed FSAR Sections 4.2.5, 4.3.1.7, 9.2.3, 9.3, 14.1.4, and 14.1.11, which corroborated the above. The inspector further reviewed the RCD Technical Manual Abnormal Operating Procedures for loss of seal injection and loss of component cooling, and found them also to be consistent with and responsive to the licensee's stated positio t No' discrepancies were identifie This item is recommended for closure L by the Resident Inspector I 18  !

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Exit Meeting An exit meeting was conducted on August 21, 1987, at which the findings of this inspection were presented to the Plant' Manager and his staf l l

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ATTACHMENT 1 PERSONS CONTACTED The following persons were contacted during and provided material input to the inspectio ,

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A. Bless Regulatory Affairs Engineer J. Carlson Technical Staff Engineer D. Clark Nuclear Training Supervisor D. Dahlen Technical Staf f Group Leader S. Kapsalis Technical Staff Engineer G. Kassner Health Physicist P. LaBlond Nuclear Licensing Assistant F. Lentine SNED - Zion M. Lesnet Technical Staff Engineer I. Netzel Project Engineer T. Reick Technical Support Superintendent S. Sanderson Licensee Contractor C. Schultz Regulatory Affairs Administrator 4 B. Soares ISI Engineer J. Tiemann Technical Staff Engineer N. Valos Operating Engineer D. Wozencraft Project Engineer

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ATTACHMENT 2

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REFERENCES AND DOCUMENTS REVIEWED Note i

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• Commonwealth Edison Correspondence

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Unless otherwise noted (by specific subject) the licensee submittals and correspondence 11sted below addressed multiple NUREG 0737 subjects and action plan items too numerous to specifically lis In such cases, the subject of such documents'is not provide .Swartz to Eisenhut,- 6/4/82 Abel to Eisenhut, 12/31/80 Reed to Denton, Response to GL 82-33, 4/14/83 Cascarano to Denton, Submittal of Procedu'res. Generation Package for E0Ps,.

12/28/84 Alexander to Denton, Response to GL 85-12,.8/22/85 i

Alexander to Denton, Response to GL 85-12, 9/15/85 Abel to Eisenhut,. 12/15/80

. Del GeorgeLt o Eisenhut, 7/1/81 Peoples to Eisenhut, 12/14/79 Abel to Eisenhut, 12/31/80 Barnes to Denton, GL 83-37, 12/6/83 Cascarano to Denton, Deviations from Containment High Range Monitor Calibration Requirements, 3/21/85 l Technical Specification Change Request, Containment Isolation (TS 3/4.9),

2/21/86 j

LeBlond to Denton, NUREG 0737, Item I.D.1, DCRDR Final Summary Report, i 5/1/86 j

'LeBlond to Denton,.NUREG 0737, Item I.D.1, DCRDR, Response to NRC Comments (Varga to Farrar letter of 8/21/86), 12/30/86 LeBlond te Denton, SPDS Human Factors Review (status), 12/31/86

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Swartz to Denton, Supplemental Response - SPDS, 7/20/83 Swartz to Denton, CECO SPDS Safety Analysis, 12/29/83 Swartz to Denton, SPDS Parameter Selection, 2/10/84 Cascarano to Denton, SPDS Parameter Selection, Display Data Validation, and Isolation, 9/7/84

l Swartz to Eisenhut, Additional Information, Item II.B.1, RCS Vents, l 5/17/82 Barnes to Denton, Response to Generic Letter 83-37 - Status of NUREG 0737 Technical Specifications, 12/6/83 Cascarano to Denton, Proposed Amendment - RCS Operational Components Technical Specification (Item II.B.1, RCS Vents)

Peoples to Eisenhut, NUREG 0578 Items 2.1.7.a and 2.1.7.b, AFW Initiation

& Indication, 12/14/79 Peoples to Denton, Containment Venting and Purging, 12/14/79 Reed to Keppler, Containment Isolation Valve Reset, 5/14/79 Technical Specification Change Request, Conditions for Purge (TS 3/4.9),

2/21/86 Lentine to Denton, NUREG 0737 Items II.F.1.4, .5, and .6, 5/13/83 Lentine to Denton, Generic Letters83-10c & d, Automatic Trip of Reactor Coolant Pumps, 4/22/83 Barnes to Denton, Generic Letters83-10c & d, Automatic Trip of Reactor Coolant Pumps, 1/4/84 Barnes to Denton, Generic Letters83-10c & d, Automatic Trip of Reactor Coolant Pumps, 3/26/84 Alexander to Denton, Response to GL 85-12, Reactor Coolant Pump Trip Criteria, 8/22/85 Alexander to Denton, Response to GL 85-12, Reactor Coolant Pump Trip Criteria, 9/18/85 Barnes to Eisenhut, Item II.K.3.25, Effect of Loss of AC on RCP Seals, 1/8/82 2. NRC Correspondence Supplement 1 to NUREG 0737 (Generic Letter 82-33), Requirements for Emergency Response Capability Varga to Farrar, Confirmatory Order, Supplement 1 to NUREG 0737, 6/12/84

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Information Notice 86-64, Deficiencies in Upgrade Programs for Plant Emergency Operating Procedures, Original and Supplement 1 Varga to Farrar, 9/9/83, Safety Evaluation Report, Item II.B.I.3, RCS Vents; II.F.1.6, Containment Hydrogen Monitor

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Thompson to Westinghouse Licensees, Generic Letter 83-10c, TMI Item II.K.3.5, Automatic Trip of Reactor Coolant Pumps, 2/3/83

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Thompson to Westinghouse Licensees, Safety Evaluation Report on !

Westinghouse Owner's Group Response to GL 83-10c and 10d, 6/28/85 Varga to Del George, Item III.K.3.25, Loss of AC on RCP Seals, 5/26/85 l Thompson to Westinghouse Licensees, Generic Letter 85-12, TMI Action Item I II.K.3.5, Automatic Trip of Reactor Coolant Pumps, 6/28/85 Varga to Del George, Confirmation of No Deviations With NUREG 0737, {

1/8/86 i Safety Evaluation Reports, Item II.E.4, Containment Isolation Dependability, 9/9/81, 9/17/81, 9/30/82 i Varga to Farrar, DCRDR Comments, 8/21/86 Varga to Farrar, DCRDR Comments, 3/10/87 Varga to Farrar, SPDS SER (Partial), 6/22/84 l Varga to Farrar, SPDS SER - Isolation Devices, 6/26/86 Varga to Del George, RCS Vents - Request for Additional Information, 3/16/82 Varga to Farrar, RCS Vents, SER - Results of Preimplementation Review, !

Item II.B.1, RCS Vents, 9/9/83 Norris to Farrar, Amendment 86 to Facility Operating Licenses (RCS Vent Technical Specifications), 2/5/85 i

Varga to Abel, Acceptability of Purge Operation, 1/15/81 }

Varga to Del George, Partial SER - Containment Purge & Vent, 9/9/81 i Varga to Del George, Completion of Review of NUREG 0737, II.E.4.2.7 and 7, 9/30/82

{ f Varga to Farrar, Demonstration of Containment Purge and Vent Valve Operability, 4/3/84 i

Varga to Del George, Status of NUREG 0737 Item II.F.1.3, High Range l Radiation Monitors, 1/8/82 I Varga to Del George, NUREG 0737 Item II.F.1, 2/8/83

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Varga to. Farrar, NUREG 0737 Item:II.F.1, 3/17/83 Varga to Del George, Acceptability of TMI Action Item II.K.3.25, Loss of AC on RCP Seals, 5/26/82 General References NUREG 0800, Standard Review Plan, 1985 Zion Updated Safety Analysis Report 10CFR50.44, Standards for Combustible Gas Control System in Light Water Cooled Power Reactors Zion Technical Specifications, Units 1 & 2 Zion System Descriptions AC Electrical Power Systems DC Electrical Power Systems Reactor Protection System Engineered Safety Features Primary Containment and Support Systems

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Westinghouse Owners Group Emergency Response Guidelines (ERGS)

. Westinghouse Owners Group ERG Executive Volume, 4/29/85 Plant Procedures, Drawings, Et I l

ST 23-81 Hydrogen Analyzer Test, Performed 12/9/81 j ZRP 1810-4 Appendix I, Shiftly Hydrogen Monitor Surveillanc ZAP 5-51-4A E0P Verification Procedure, Revision 2

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Completed Attachments A thru D for verification of E-3  ;

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Series E0Ps (listed below)

ZAP 5-51-4B E0P Validation Procedure, Revision 1 l Completed validation Test Scenarios (10 examples, completed 3/85 - 12/85)  ;

Completed Table Top Validation - ES ZAP 5-51-5 Procedure Content and Format, P ,ision 28 ZAP 5-51-5A E0P Content and Format, Revision 1 E-0 Series Emergency Operating Procedures - Reactor Trip or Safety Injection, 12/27/85 E-1 Series Emergency Operating Procedures - Loss of Reactor or Secondary Coolant, 12/22 & 27/85

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E-3 Series Emergency.0perating Procedures - Steam Generator

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Tube Rupture (Including ~ Emergency Subprocedures, Emergency it '

. Contingency Action Procedures, Functional Restoration Procedures'and Status Trees), 12/27/85 ZAP 10-52- Radchem Foreman Shift Turnover Sheet,' Revision 0

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S&L Piping Design Table E-1, Revised.12/30/70 2A-RV147 Containment Hydrogen Monitor Calibration, Rev. O M52/515- Reactor Coolant System P&ID I M53/516 Reactor Coolant System P&ID

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M25/505 Condensate System P&ID M70/70 Containment Air Monitoring P&ID I

M54/517 Chemical and Volume' Control P&ID M55/518 Chemical and Volume Control P&ID

'M56/519 Chemical and Volume Control P&ID

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M64/521 ' Safety Injection System P&ID M65/522 Safety Injection System P&ID

.M76/535 Containment HVAC P&ID i

M66/523 Component. Cooling P&ID M956/959 Reactor Vessel Head Vent P&ID i

M22/502 Steam Generator Feedwater M499 Radiation Monitor Location & Support Detail l

.22E-1-3000 Electrical Diagrams - Radiation Monitor 22E-1/2-4000B 4KV One Line Diagram 22E-1/2-4000C 4160 & 480 V ESS Busses One Line Diagram 22E-1/2-4000H 4160 & 480 V ESS Busses One Line Diagram, Bus 47 &

22E-1/2-4000J 4160 & 480 V ESS Busses One Line Diagram, Bus 49 22t'-1-4525 Electrical Diagram, ESFAS 22E-1-4635 Electrical Diagram, ESFAS

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22E-1-4636 Electrical Diagram, ESFAS

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22E-1-4637 Electrical Diagram, ESFA _

- 22 E-1-4913 -' Electrical Diagram, ESFAS

thru 4923 22E-1-4929- Electr.ical Diagram,-.ESFAS 22E-1-4915 Electrical Diagram, ESFAS

'22E-1-4916

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Electrical Diagram, ESFAS

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22E-1-4840 Containment P&E Valve Schematics,~Shs RV 43 - 46 i

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22E-1-4840 Containment Isolation Valve ~ Schematics, j Shs RV63 - 70;.FW2,3; DT3, DT96, SS18

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565D30 Westinghouse Functional Diagrams - ESFAS, Shs 1-16 IL-519T Calibration SG'A Level Transmitter Revision 3, 5/20/87-2 L-503 20 SG Wide Range Level Calibration, Revision 2, 5/21/87

'2FW-FWO2 Aux FW Flow Loop 1 Calibration, Revision .1' ,

4/27/87 22E-2-4840 AFW Pump 2B Control Elementary Diagram, Sh FW2, ;

Rev. AE 22E-2-4840 AFW Pump 2A Control Elementary Diagram, Sh FW3, Rev. AE 22E-2-4840 AFW Pump 2A Control Elementary Diagram, Sh FW 58, ;

Rev. T {

TSSP 96/96-85 GA High Range Rad Monitor Special Calibration, Rev. 0 1 E-255-978 GA Energy Response Test & Dose Rate Calibration of ']

Model RD23 High Range Radiation Monitor Detector, 5/81 )

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ECA 281-9034 Test Procedure for High Range Radiation Monitoring )

System, 9/25/80 & 7/11/80 M22-2-79-55 Modification-Package, High Range Containment Monitor, 2/7/83 ST 2-82 Functional Test - Containment High Range Rad

, Monitor, Revision 1

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PT-7 Auxiliary Feedwater Pumps Tests and Checks, 7/3/87 PT-7A Starting Procedure for Motor Driven Aux FW Pump Lube Oil Pumps, 6-17-87 PT-10 U1/02 Safeguards Actuation Tests, 4/6/87 -l PT-10A/B iJ1/U2 Safeguards Logic Tests, 4/21/87 295-103- Action Item Report, Licensee Evaluation of '

!86-064 Information Notice 86-64 Memo Abel to CECO Staff, Verification and Validation of SPDS l Display Systems, 6/6/84

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SPDS Design Description, Rev. Safety Parameter Display System Verification and Validation Program & Report, 7/8/87

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SPDS Requirements Document, CYGNA Corp., Rev. 2

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SPDS Review Guide / Checklist, Revision 1

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SPDS Human Factors Design Checklist & Results

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SPDS Operations Design Review Checklist & Results l

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SPDS Nuclear Safety Design Review Checklist & Results !

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SPDS - Resolution of Design Review Open Items, 5/87

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Offsite Review 83-21, SPDS, and response, Abel letter to Del George, 12/28/83 SNED Pro Systems Integration Checklist, Revision 15, performed 4/22/87 for SPDS Memo Kaptur to Sorrentino, SPDS Alarm Discrepancies &

Subcooling Calculation Comments, 3/15/83 M22-1/2-79-43 Reactor Vessel Head Vent Modification Packages, Units 1 & 2 l

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Preop Test RV Head Vent Preoperational Valve Limit Switch l Functional Check, 4/22/81  ;

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. Preop Test RV Head Vent Preoperational Valve Functional Test Under Pressure, 2/11/81 Preop Test RV Head Vent Hydrostatic Test, 3/30/83 i

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e f , . Preop. Tes RV-Head Vent Solenoid Valve Functional Test 3/3/81 ,

I' LTSS;15.6 20V-P. Technical Staff Surveillance Procedure, RV Head

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Flow Test,.2/87 & 2/87' ,

TSS 15.6.103' LTechnical Staff Surveillance Procedure, Reactor 1 Head Vent Operability Test, Revision PT-27z Miscellaneous Valve Test, Revision 13 (RV Head Vent)

501-9 Containment-Ventilation System Operating Instruction, 5/18/87 j S01-10 Auxiliary Feedwater, '.5/18/87 S01-62 Radiation Monitoring, 3/11/86

A0P- High Radiation Alarm, 6/17/87 )

I A0P-lil Excessive RCS Leakage, 6/17/87 J

A0P-4.2' ' Inadequate AFW Supply 6/17/87-A0P-1. 2 :- Steam Generator Tube Leak, 6/17/87

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A0P- RCP Malfunction, 6/17/87 l

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A0P-4.1' Loss.of Component Cooling Water, 6/17/87

~A0P- Operation with a Failed Instrument Channel,

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6/17/87 A0P- Control Room Inaccessibility, 6/17/87  !

Panel 3-2E Annunciator Response - Containment Purge Valve ,

ESS Bypass, 7/18/81 '

Panel 3-5A . Annunciator Response - Ventilation Isolation, j 3/20/86 l Panel 3-5B Annunciator Response - CI-A, 3/23/86

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Panel 3-5C Annunciator Response - CI-8, 7/18/81 Panel 3-6A Annunciator Response - Area Monitor Radiation High, 2/11/82 j

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1987 Licensed Operator Requalification Training Schedule, Rev. 1, 2/14/87

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Licensed Operator Retraining Lesson Plan, E0P E-1 and Related Procedures, Rev. 0

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Licensed Operator. Retraining Attendance Records, 11/82, 1-4/84, 10-11/84, 3-4/87

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Licensed Operator." Current Year Plant Mod" Required Reading, 1983 & Related Exam #T71-1

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Licensed Operator Training Lesson Plan i

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LO-PSC-40, Plant Process Computer System (SPDS), 1/28/87 M22-1/2-79-55 Modification Package, High Range Radiation Monitor, 2/7/83 TSSP-95/96-85 GA High Range RM Special Calibration Procedures, Revision 0, 5/9/85, 12/3/85. 12485, 12/5/85 E-255-978 GA Energy Response Test & Dose Rate Calibration of Modem RD-23 High Range. Radiation Monitor Detector, 5/81 EL-281-9034 Test Procedure for High Range Radiation Monitoring System, Revision A, 9/25/80 & 7/11/80 ,

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ST-2-82 Functional Test - Containment High Range Monitor, ,

Revision 1

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Environmental Qualification' Data Sheets, High Range l Radiation Monitors 1 REAR 02 & 03, 2 REAR 02 & 03 i i

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Environmental Qualification Report No. EQ-ZN023 for High 1 Range Radiation Monitors i M22-1/2-73-39 Installation of Hydrogen Monitors i

Preop Test H2 Monitor Containment Isolation Valves, 9/15/81 Preop Test H2 Monitor Leak Test, 12/21/81 ZRP-1810-4 H2 Analyzer Functional Test, Rev. O, 4/14/87 4/20/87, 4/7/87, 3/23/87, 3/16/87, 3/2/87, 2/19/87, 2/5/87 2A-RV-147 Containment H2 Monitor Calibration, Rev. O  !

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Comsip Delphi Drawing 1111, H2 Analyzer Schematic, Rev. 1 l l

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Letter, Comsip Delphi to CEC 0, Containment Hydrogen l Monitor Alarm Wiring Modifications, 4/1/82  ;

R22-1/2-87-55 Modification Request Form, Containment Sampling l Piping Inadequately Supported, 7/8/87

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REGION III ITEM CLOSURE FORM FOR Zion Units 1 & 2 ITEM NUMBER : INTERIM : INDIVIDUAL : CLOSURE :

DKT/YY-XX-XX : REPORT N0 : ASSIGNED ITEM : REPORT NO. :

(1) : (95) : (106) : (121) :

295/86013-04 :  :  : 295/87028 :

304/8601/-05 :  : Holzmer : 304/87029 :

P :  :  :  :

295/86013-05 :  :  : 295/87028 :

304/86012-06 :  : Holzmer : 304/87029 :

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295/85005-06 :  :  : 295/87028 :

304/85005-06 :  : Holzmer : 304/87029 :

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COMPLETED BY:_ Holzmer PAGE 10F 1

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i REGION III' TRACKING SYSTEM

~ Zion Units 1 & 2 SITE NEW ENTRY .__X_ MODIFY

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ITEM NUMBER (1/1): 295/86013-03 (DKT/YYXXX-XX-X)  : 304/86012-04 PERSON WHO IDENTIFIED (1/16):

FOLLOWUP DUE DATE (1/38):

{YY-MM-DD)

MODULE N (1/94):

RESPONSE DUE DATE (1/100):

(YY-MM-DD)

PERSON ASSIGNED (1/106): Holzmer i CLOSE0VT REPORT N (1/121):

(DKT/YYXXX)

INTERIM REPORT N (2/95): 295/87028 (DKT/YYXXX) 304/87029 INSPECTOR (2/106): Beckman i BRIEF DESCRIPTION:(31)

RCS vent not environmentally qualifie Memo to HQ (RTA).__:D0 NOT TYPE PAST drafted and sent to Region III 10/2/87 :THIS POINT

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REGION III TRACKING SYSTEM Zion Units 1 & 2 SITE NEW ENTRY X MODIFY

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l ITEM NUMBER (1/1): 295/85031-02 (DKT/YYXXX-XX-X) : 304/85032-04 PERSON WHO IDENTIFIED (1/16):  ;

FOLLOWUP DUE DATE (1/38):

(YY-MM-DD)

MODULE N (1/94):

RESPONSE DUE DATE (1/100):

.(YY-MM-DD1 PERSON AS*?IGNED (1/106): Holzmer CLOSE0VT REPORT N (1/121):

(DKT/YYXXX)

INTERIM REPORT N (2/95): 295/87028 (DKT/YYXXX) 304/87029 INSPECTOR (2/106): Beckman i

BRIEF DESCRIPTION:(31)

Conflict between Zion TS & Confirmatory Order item :00 NOT TYPE PAST E.1.c. Licensee will attempt to resolve by having Order :THIS POINT Item E.1.c removed based on TS Amendment 80/7 :

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I REGION III TRACKING SYSTEM Zion Units 1 & 2 SITE

_X_[NEWENTRY - MODIFY

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ITEM NUMBER (1/1): 295/87028-01-0 (DKT/YYXXX-XX-X)  : 304/87029-01-0 PERSON WHO IDENTIFIED (1/16): Beckman FOLLOWUP DUE DATE (1/38):

(YY-MM-DD)

MODULE N (1/94): 25565 RESPONSE DUE DATE (1/100):

,(YY-MM-DD)

PERSON ASSIGNED (1/106): Holzmer CLOSE0VT REPORT N (1/121):

(DKT/YYXXX)

INTERIM REPORT N (2/95):

~

[0KT/YYXXX)

INSPECTOR (2/106):

BRIEFDESCRIPTION:(31)

TMI item II.F. Hydrogen sample piping not secure :00 NOT TYPE PAST Open pending RIII review of CECO corrective actio :THIS POINT

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REGION III TRACKING SYSTEM l'

X ~NEW Zion ENTRY Units 1 -&MODIFY 2 SITE

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ITEM NUMBER (1/1): 295/87028-02-0 (DKT/Y'Y XXX-XX-X ) : 304/87029-02-U PERSON WHO IDENTIFIED (1/16): Beckman FOLLOWUP DUE DATE (1/38):

(YY-MM-DD)

MODULE N (1/94): 25565 RESPONSE DUE DATE (1/100):

.(YY-MM-DD)

PERSON ASSIGNED (1/106): Holzmer '

CLOSE0VT REPORT N (1/121):

[DKT/YYXXX)

INTERIM REPORT N (2/95):

[DKT/YYXXX)

INSPECTOR (2/106):

3,

'J BRIEF DESCRIPTION:(31)

TMI item II.K.3.5. LicenseeffailedtorespondtoGL85-12 :00 NOT TYPE PAST item A-2 as committed in thdr 8/22/85 letter. Also :THIS POINT modification to replace Barton with Rosemount. Open  : ,

pending review of licensee's submittal of uncertainty :

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analysis for Norm / Adverse containment for RCS WR Pressure _ :

Submittal expected by Mid November, 1987  :

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