IR 05000277/1987009

Insp Repts 50-277/87-09 & 50-278/87-09 on 870314-23 & 0410-24.No Violations Noted.Major Areas Inspected: Operational Safety,Radiation Protection,Physical Security, Control Room Activities,Maint & Outstanding Items

ML20214S136
Person / Time
Site:	Peach Bottom
Issue date:	05/26/1987
From:	Gallo R, Linville J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20214S113	List: IR 05000277/1987009 ML20214S110
References
50-277-87-09, 50-277-87-9, 50-278-87-09, 50-278-87-9, NUDOCS 8706090150
Download: ML20214S136 (24)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION I

Report Nos. 50-277/87-09 & 50-278/87-09 Docket Nos. 50-277 & 50-278

License Nos. DPR-44 & DPR-56 Licensee: Philadelphia Electric Company 2301 Market Street

Philadelphia, Pennsylvania 19101 Facility Name: Peach Bottom Atomic Power Station Units 2 and 3 Inspection At: Delta, Pennsylvania-Inspection Conducted: March 14 to 23, 1987 and April 10 to 24, 1987 Inspectors: T. P. Johnson, Senior Resident Inspector R. J. Urban, Re ident I pector i

Reviewed By: M , 2- .

& F7 J.g UfMTie', CN ( date /

F4 tor Pr ts ction 2A C.v sian o e tor P o ects Approved By: k, r 5'lMl87 R. M7Gallo, Chief date i Projects Branch No. 2 Division of Reactor Projects Inspection Summary: Routine, on-site regular and backshift and weekend resident inspection (82 hours9.490741e-4 days <br />0.0228 hours <br />1.35582e-4 weeks <br />3.1201e-5 months <br /> Unit 2; 81 hours9.375e-4 days <br />0.0225 hours <br />1.339286e-4 weeks <br />3.08205e-5 months <br /> Unit 3) of accessible portions of Unit 2 and 3, operational safety, radiation protection, physical security, control room activities, licensee events, surveillance testing, refueling and outage activities, maintenance, and outstanding item Results: The inspector identified two unresolved items: local leak rate

testing and control room upgrade modification. One instance of inadequate j shift turnover and associated logkeeping was noted.

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8706090150 870529

{DR ADOCK 05000277

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DETAILS Persons Contacted B. L. Clark, Administrative Engineer

• J. B. Cotton, Superintendent Plant Services G. F. Dawson, Maintenance Engineer

• A. B. Donnel, QA Site Supervisor

• R. S. Fleischmann, Manager, Peach Bottom Atomic Power Station A. A. Fulvio, Technical Engineer D. P. Potocik, Senior Health Physicist J. F. Mitman, Radwaste Engineer R. H. Moore, Superintendent, Quality Assurance Division D. L. Oltmans, Senior Chemist F. W. Polaski, Operations Engineer

• D. C. Smith, Superintendent Operations J. E. Winzenried, Staff Engineer Other licensee employees were also contacte *Present at exit interview on site and for summation of preliminary finding . Plant Status 2.1 Unit 2 Unit 2 began the inspection period in a refueling outage that started on March 14, 198 Fuel bundle #49-40, which stuck to its fuel support piece, was removed from the core on April 17, 1987 (see section 4.4.3). The core off-load was completed on April 19, 198 On April 22, 1987, a new control rod blade was dropped in the spent fuel pool (see section 4.4.4). At the end of the inspection period, control rod blade shuffles were in progres .2 Unit 3 The unit began the inspection period reducing load on March 14, 1987, for a control rod pattern adjustment. On March 17, 1987, the unit scrammed from 85% reactor power due to EHC problems (see NRC Inspection 277/87-10; 278/87-10). The unit was restarted on March 21, 1987. The unit scrammed from 1% reactor power on March 25, 1987, on low reactor water level (see NRC Inspection 277/87-10 278/87-10). t The unit restarted on March 27, 1987, and full reactor power was achieved March 31, 198 .

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The unit was shutdown due to an NRC Order (see section 4.2) on March 31, 198 Cold shutdown was achieved on April 1, 198 The unit remained in cold shutdown for the remainder of the inspection perio . Previous Inspection Item Update 3.1 (Closed) Inspector Follow Item (277/86-03-01). RPS Power Supply Breakers Long Term Corrective Actions. The licensee's short term and long range corrective actions were previously reviewed in NRC Inspection 277/86-13 and 278/86-14. Modification (MOD) 1916 added a series circuit interlock to each RPS shunt trip coil circuit to ensure that the shunt trip coil is not energized beyond its momentary ratin The licensee reported the completion of MOD 1916 in a letter dated March 19, 1987, for Unit The inspector reviewed the letter, MOD 1916 documentation, and discussed this item with licensee engineers. Based on the above, the inspector follow item is close .2 (Closed) IE Bulletin No. 86-03 (277/86-8U-03; 278/86-BU-03). ECCS Pump Failures Due to Single Valve Failures. The licensee responded to the Bulletin in a letter dated November 10, 1986. The inspector reviewed the licensee's response and appropriate electrical schematic and piping diagrams. The licensee determined that the single failure problem does not exist with the Peach Bottom ECCS pump minimum flow valve design. This is based on each pump having an individual motor operated valve (MOV) powered from different electrical division Also, the MOVs actuate to open on a low pump flow condition. Each pump and MOV are physically separated from the other redundant pumps and MOVs. Based on the licensee's response and inspector's the review, lE Bulletin No. 86-03 is closed for both Unit 2 and . Plant Operations Review 4.1 Station Tours The inspector observed plant operations during daily facility tours. The following areas were inspected:

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Control Room

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Cable Spreading Room

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Switchgear and Battery Rooms

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Reactor Buildings

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Turbine Buildings

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Radwaste Building

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Recombiner Building

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Pump House

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Diesel Generator Building

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Protected and Vital Areas

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Security Facilities (CAS, SAS, Access Control, Aux SAS)

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High Radiation and Contamination Control Areas

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Shift Turnover

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4. Control Room and facility shift staffing was frequently checked fcr compliance with 10 CFR 50.54 and Technical Specifications. The presence of a senior licensed operator in the control room was verified frequentl . The inspector frequently observed that selected control room instrumentation confirmed that instruments were operable and indicated values were within Technical Specification requirements and normal operating limit ECCS switch positioning and valve lineups were verified based on control room indicators and plant observation Observations included flow setpoints, breaker positioning, PCIS status, and radiation monitoring instrument . Selected control room off-normal alarms (annunciators)

were discussed with control room operators and shift supervision to assure they were knowledgeable of alarm status, plant conditions, and that corrective action, if required, was being taken. In addition, the applicable alarm cards were checked for accuracy. The operators were knowledgeable of alarm status and plant condition With Unit 2 shutdown for refueling, the inspector noted that the Unit 2 division 1, 2A/2C 125 volt DC battery ground annunciator window was alarmed in the control room on March 23, 1987. The inspector reviewed the alarm response card #2(3)0C209R-26 and discussed the required

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actions with the on shift operators. The actions require a non-licensed operator to check battery ground lights in the cable spreading room and to notify the shift superviso Procedure S.8.5.G, " Procedure for Investigating DC Battery Grounds", Revision 1, provides for checks for the condition and provides actions to determine the source of the groun The licensee had implemented procedure S.8.5.G and had determined that the source of the ground was not the annunciator DC feed FSAR section 8.7.4.2 states that the DC battery systems are designed so that the only reasonable failure that can be postulated is a multiple ground which would interrupt only the grounded circuits. The probability of this occurrence is very small because the first ground would be detected by a ground alarm, and would then be located and corrected. Battery conditions are observed by regular checking for any deterioration. Low battery voltage is also annunciated in the control roo _ - - .. . . _

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i The licensee has a method to determine non-annunciator ground conditions by utilizing a " battery eliminator" device. The battery eliminator provides an alternate source to the DC loads, thereby allowing the ability to check for grounds with installed ground detection circuitry in the cable spreading room. This prevents interrupting DC power to vital loads. The licensee has drafted a procedure for the use of a battery eliminator, however it has not yet been PORC approved. Apparently, the licensee is reviewing the seismic and safety related qualifications for the battery eliminator. (NRC Inspection 277/86-25 also addresses this item.)

The inspector discussed DC grounds, procedure S.8.5.G, FSAR Section 8.7.4.2 and the proposed battery eliminator procedure with licensee engineers and operators. In addition, the inspector reviewed the appropriate electrical schematic drawings for the battery systems and ground detection circuitry. The inspector will review the proposed battery eliminator procedure in a future inspectio No violations were note .1.4 The inspector checked for fluid leaks by observing sump status, alarms, and pump-out rates; and discussed reactor coolant system leakage with licensee personne .1.5 Shift relief and turnover activities were monitored daily, including backshift observations, to ensure compliance with administrative procedures and regulttory guidanc On a control room tour at 7:40 a.m. on April 15, 1987, the inspector noted that Unit 2 control rod 34-23 indicated that the rod was at position 10 on the full core display. An 00-7 option 2 printout at 6:57 a.m. also indicated the rod was at position 10. The inspector noted that the Unit 2 licensed reactor operator was pursuing this information. Unit 2 was partially defueled at the time and the fuel surrounding control rod 34-23 had been remove Previous 00-7 option 2 printouts at 8:08 p.m. and 11:05 ,

p.m. on April 14, 1987, indicated that rod 34-23 was at '

position 1 The licensee independently sent two senior reactor operators (SR0s) to the fuel floor to make a visual check of the control rod position. Both SR0s visually determined that control rod 34-23 was fully inserted to position 00. The inspector confirmed this information through discussions with each SR _ _ _ . . - .

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Apparently, individuals doing an under vessel inspection on "Y" shift (3:00 p.m. - 11:00 p.m.) on April 14, 1987,

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bumped the rod position indication probe and it became i disconnected, thus giving an erroneous indication. The

"Y" and "Z" shifts had discussed and pursued this

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problem, and apparently concluded that the control rod was fully inserted. However, this information was neither logged nor documented on the shift turnover sheets, nor passed on verbally to the "X" shift (7:00 a.m. - 3:00 p.m.) R Licensee operations management counselled the reactor operators involved, and issued a memo that was read to each operating shift. The inspector reviewed the April

-; 20, 1987 memo. The inspector discussed this item with  ;

licensed operators and licensee management personne The inspector also monitored the "Y" shift turnover

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meeting on April 23, 1987, at which this information was

{ discussed. The inspector stated that this was an

, example of incomplete logkeeping and incomplete shift I turnover. The inspector has no further questions at

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this time. Shift turnover will be reviewed in future ,

inspections.

4. The inspector observed the main stack and both reactor i

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I building ventilation stack radiation monitors and recorders, and periodically reviewed traces from backshift periods to verify that radioactive gas release rates were

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i 4. The inspector observed control room indications of fire detection instrumentation and fire suppression systems, l monitored use of fire watches and ignition source controls, checked a sampling of fire barriers for integrity, and j observed fire-fighting equipment stations. No inadequacies

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4. The inspector observed overall facility housekeeping .

conditions, including control of combustibles, loose *

trash and debris. Cleanup was spot-checked during and j after maintenance. Plant housekeeping was generally j acceptable.

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4. The inspector observed the nuclear instrumentation

subsystems (source range, intermediate range and power r i range monitors) and the reactor protection system to

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verify that the required channels were operable.

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4.1.10 The inspector frequently verified that the required off-site electrical power startup sources and emergency on-site diesel generators were operabl .1.11 The inspector monitored the frequency of in plant and control room tours by plant and corporate managemen The tours were generally adequat .1.12 The inspector verified operability of selected safety related equipment and systems by in plant checks of valve positioning, control of locked valves, power supply availability, operating procedures, plant draw-ings, instrumentation and breaker positioning. Selected major components were visually inspected for leakage, proper lubrication, cooling water supply, operating air supply, and general conditions. No significant piping vibration was detected. The inspector reviewed selected blocking permits (tagouts) for conformance to licensee procedures. Systems checked included the Standby Gas Treatment System. No inadequacies were identifie .1.13 The inspectors performed backshift and weekend tours of the facility on the following days:

"Y" shif t (3:00 p.m. - 11:00 p.m.) - March 19, 20, 23; April 14, 16, 17, 20, 22

"Z" shift (11:00 p.m. - 7:00 a.m.) - March 17, 19, 20, 23; April 13, 14, 16, 2 .2 NRC Order Suspending Power Operation Dated March 31, 1987 4. PECo Response

. In a letter dated April 6, 1987, the licensee responded to the NRC Order. The response was required within seven days and includes the following items / actions:

4.2. Establishment of a 24-hour coverage of operations by at least one independent, nuclear experienced engineer or physicist per shift who will be posted within the control room complex to observe licensed duties and report directly to the Super-intendent of Nuclear Operations Quality Assurance Division at the Corporate Headquarters. If it becomes necessary for this individual to leave the control room complex for a brief period,

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their functions will be assumed by the Shift Technical Advisor during their absence. The function of these individuals will be to verify

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that licensed personnel are alert and profes-sional in the manner they conduct licensed duties in accordance with NRC requirements and station procedures. This 24-hour coverage was commenced April 10, 198 .2. Immediately after receipt of the Order, Senior Plant Management began making unannounced visits to the control room during each night shift (11:00 p.m to 7:00 a.m.). The activity continued until the 24-hour coverage of operations by the special monitoring team was established, at which time the frequency of these visits was reduce .2.1.3 An administrative block (tagout), which ensures that the shutdown conditions are maintained, was applied on March 31, 1987 at 5:30 p.m. The block requires the mode switch of each unit to be in the SHUTOOWN or REFUEL position to ensure com-pliance with the Order. Removal of the block requires the approval of the Plant Manager or the Superintendent-Operation .2.1.4 Meetings chaired by the Plant Manager were held with control room operating shift personnel, beginning on March 31, 1987, to discuss the Order and events leading to the Order. During these meetings, it was forcefully restated that sleep-ing or the appearance of sleeping at their posts is unacceptable and will result in immediate suspension from duty with recommendation for termination of employment. Personnel were urged to be open, frank, and candid during interviews with the NRC on this subjec .2.1.5 Special meetings, in addition to the above meet-ings, were held with Shift Superintendents and Shift Supervisors (senior licensed operators).

In these meetings Shift Supervision was force-fully reminded by the Plant Manager that it was their responsibility to ensure that personnel on their shift remain attentive and alert, and to take appropriate action if this is not the cas Furthermore, they were informed that if they received reports of such situations concerning other shifts or individuals that those reports should be forwarded to Senior Plant Management for dispositio .

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4.2. Upon receipt, the Order was posted at the station and a summary was issued via a station newsletter. Subsequently, a copy of the Order was distributed to all control room personne .2. The Peach Bottom Operations Engineer was reassigned to Corporate Headquarters effective April 6, 1987 and has been replaced with another nuclear experienced, qualified engineer holding a senior operator license at the Peach Bottom statio .2. The following human factor related changes were made to the control room to make the operators more observable by the Shift Supervisor: The Shift Supervisor's post in the control room has been elevated to give him a better view of the room and the operators. Additionally, the high-back chairs in the control room have been replaced with low-back chair .2. Control room operators are being required to record certain key plant parameters and the decay heat removal equipment status hourly with review and sign-off by shift supervision to ensure that the cold condition is safely maintaine The recorded data includes: mode switch position, reactor water level and temperature, reactor pressure, reactor head vent valves position (not applicable if reactor head is removed), decay heat removal equipment status (in service or standby), and fuel pool temperatur .2.1.10 An additional status report to Senior Plant Management is now required of the afternoon ("Y")and("Z")nightshift Shortly prior to the end of these shifts, Shift Supervision is being required to report to Senior Plant Management the plant status and a brief synopsis of his shift's activitie .2.1.11 All licensed control room operators are being required to re review administrative procedure A-7, " Shift Operations", and IE Circular 81-02,

" Performance of NRC Licensed Individuals While on Duty", and sign a statement documenting that they have read and understand the materia _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ . _ _ - _ _ _ - _ - . _,

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The inspector verified that the actions committed to in F-l section 4.2.1 above were being implemente .

I The inspector reviewed the Nuclear Operations Monitoring i Team (NOMT) implementation including a review of the

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April 7, 1987, Nuclear Operations QA Division memo establishing NOMT

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Memo on the organization, duties, and administration of NOMT (no date)

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Checklist for Independent Observation of Shift l Operation

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April 6, 1987, letter from John S. Kemper to T. ! Murley in response to the NRC Order i

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J. L. Everett, Chairman and CEO News Conference

dated April 6, 1987 a

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IE Circular 81-02: Performance of NRC-Licensed j Individuals While on Duty

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Shift lineup schedule for NOMT

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List of NOMT personnel and schedule i --

NRC Order dated March 31, 1987 l --

Memo from J. W. Gallagher dated March 27, 1987, j dealing with sleeping on the job

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Questions and Answers for NOMT, dated April 7, 1987 i'

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Drew Smith Memo " Monitoring in Control Room", dated

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April 7, 1987

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PEco Press Release dated April 6, 1987, in response 4 to NRC Order

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Administrative Procedure A-7, " Shift Operations",

j Rev. 22, dated October 8, 1986

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Procedure NOMT A-100, " Procedure for the Conduct of

Control Room Monitoring by the NOMT" Rev. O and l Rev. 1 l

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NOMT Logbook in the Control Room

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NOMT Daily Reports The inspector verified that hourly checks of unit parameters for cold shutdown are being performed per ST-9.32-2 and 3 (see section 7). The inspector verified that the reactor mode switches were blocked in " refuel" or " shutdown" per shif t block #2-87-393 dated March 31, 198 The inspector interviewed each of the NOMT personnel and their supervisor, the Superintendent. Nuclear Operations Quality Assurance Division. The NOMT personnel include one Limerick individual, one corporate individual, one Peach Bottom QC individual, the Peach Bottom ISEG supervisor, and a Peach Bottom test engineer. The first four individuals reported previously to organizations not within the Peach Bottom operations line organization. However, the Peach Bottom test engineer does report through the Peach Bottom line organization. It would be possible for the test engineer to return to his previous organization during utility shifts or at other time Thus, the independence of the NOMT may be compromised. When questioned about this, the licensee replaced the Peach Bottom test engineer with another individual who was independent of the operations organization, effective April 24, 198 The inspector will continue to periodically monitor the Itcensee commitments to the NRC Orde .3 Lois and Records The inspector reviewed logs and records for accuracy, complete-ness, abnormal conditions, significant operating changes and trends, required entries, operating and night order propriety, correct equipment and lock-out status, jumper log validity, conformance to Limiting Conditions for Operations, and proper reporting. The following logs and records were reviewed: Shift Supervision Log, Reactor Engineering Logs, Unit 2 Reactor Operator's Log, Unit 3 Reactor Operator's Log, Control Operator Log Book, NOMT Log Book, and STA Log Book, Night Orders, Radiation Work Permits, Locked Valve Log, Maintenance Request Forms Temporary Circuit Mod-ification Log, and Ignition Source Control Checklists. Control Room logs were compared against Administrative Procedure A-7, Shift Operations. Frequent initialing of entries by licensed operators, shift supervision, and licensee on-site management constituted evidence of licensee review. The Unit 2 reactor operator's log was noted as incomplete on one occasion (see section 4.1.5).

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4.4 Unit 2 Refueling Outage Activities On March 13, 1987, Unit 2 was shut down for a scheduled 71 day refueling outage. Major work items include refueling,10 CFR 50 Appendix R modifications,10 CFR 50 Appendix J testing, turbine maintenance, plant modifications, and other maintenance and i testin . CFR 50 Appendix J Testing i

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In accordance with 10 CFR 50 Appendix J, the licensee '

initiated local leak rate testing (LLRT) of the primary ;

containment inboard and outboard main steam line drain

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valves (M0-74 and M0-77, respectively). On March 15,

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between M0-74 and MO-77 in order to perform an LLR When the line could not be pressurized, maintenance request forms (MRFs) were written to repair MO-74 and M0-7 r On March 16, 1987, the licensee determined that as found leakage rates for M0-74 and MO-77 should be measured prior to maintenance. During the LLRT, the piping between both valves could not be drained. Apparently, when the vessel was flooded up for refueling, water was

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leaking by the main steam line plugs and by M0-7 MO-74 was considered unacceptable and will be repaired or plugged in order to allow testing of MO-7 Sometime during the next two to three weeks, (March 17 to April 4, 1987) M0-74 was " blue tested" by maintenance.

. The test determined that the valve seat was acceptabl The licensee determined that if MO-74 now passed an LLRT, the results could not be accepted since the valve was disturbed, even though it was not actually repaired. Also ,

during this time period, preventive maintenance was performed on the breaker for M0-77.

On April 4, 1987, the Itconsee attempted to LLRT the

! main steam line drain valves. MO-74 and MO-77 were

found open. The control room operator was requested to close both valves. M0-74 was closed but when the hand switch for M0-77 was turned to "close", the valve did not reposition. A second attempt was also made with no l valve movement. Thermal overloads in the breaker were !

l reset and the valve was partially closed using the manual handwheel. The hand switch was turned to "close" and the valve fu)1y opened and backseated. Another stroke was performed with similar results. M0-77 was then closed manually and an LLRT was performed with successful results. However, M0-77 was not electrically closed, so the test was not acceptabl _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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The licensee began additional troubleshooting of MO-7 M0-77 was manually partially opened and the hand switch was turned to "open". The valve closed and tripped on overload. Apparently, the hand switch was working backwards, and the open and close torque switches were not operable. The breaker was examined and the licensee determined that it was rewired incorrectly during preventive maintenanc On April 5, 1987, the breaker was rewired correctly and M0-77 was electrically closed in preparation for the LLRT. The leak rate was offscale on the LLRT test equip-ment, the pipe volume could not be pressurized, and the leakage path could not be determined because the pipe was cut downstream of M0-77 and two valves between MO-77 and the pipe cut were blocked open. On April 6, 1987, valves 310 and 311 were plugged. A retest was performed on April 7, 1987. The integrated leak rate test (ILRT) rotometer was used because its scale is about a factor of ten greater than the scale on the LLRT bo On April 7, 1987, at 12:30 p.m., MO-77 was found to be leaking greater than the maximum allowed leakage rat Therefore, a leakage path apparently existed in primary containment in excess of the allowable leakage rate because both M0-74 and MO-77 failed LLRTs. However, when the MO-74 valve was manually opened 1/2 turn, the leakage stoppe On April 8, 1987, at 11:00 a.m., the licensee made a four hour report in accordance with 10 CFR 50.72. The licensee stated that the report was 18-1/2 hours late because of uncertainties associated with interpreting the results of the LLRTs of MO-74 and MO-7 In summary, the Itcensee determined that MO-74 had no as found leakage data because the valve was disassembled before an LLRT could be performed. MO-77 as found leakage was greater than La. However, M0-77 was most Itkely a good valve with respect to leakage at shutdown because it served as a boundary valve during MSIV LLRTs and air makeup to maintain pressure was minimal. MO-77 was most likely damaged during troubleshooting when the valve was wired backward As part of this review, the inspector had discussions with the Shif t Technical Advisor (STA) who had made the four hour report, a test engineer who was responsible for MO-74 and MO-77 LLRis, and the Performance

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Supervisor. The inspector also reviewed Surveillance Test (ST) 20.029, "LLRT-Main Steamline Drain", Rev. 4, dated 3/21/86, conducted on March 15, 16, and April 4, 5, 7, 198 The licensee intends to submit an LER for this event. The inspector will review the LER in a future inspection. The discrepancies identified by the licensee with regards to leak rate as found testing, preventive maintenance on M0-77 and the late reportability are collectively unresolved and will be reviewed in a future inspection (UNR-277/87-09-01).

NRC Inspection 277/87-12; 278/87-14 also reviewed this item, 4. Refueling Modifications The inspector reviewed selected modifications (M00s) that are being performed during the Unit 2 outage. Items checked include MOD scope, safety evaluation, MOD check-Itsts, construction job memos, QC signoffs, PORC approval forms, installation and modification acceptance testing, infield work, maintenance request forms and permits (tagouts).

4.4. MOD 1419 replaces the safety related Rosemount model 1151 transmitters for reactor pressure, main steam line differential pressure, and reactor water level instruments. The inspector reviewed the pneumatic pressure test requirements that checks the instrument piping integrity per ANSI /ASME B31.1-1980 code, section 13 The licensee is performing a 100 psi pneumatic leak check of the tubing for 10 minutes. The inspec-tor verified this by reviewing test documentation, and discussed it with licensee personnel and NRC regional specialists. No violations were note .4. MOD 1729 remodels the Control Room to optimize the working environment by implementing human factors recommendations. Changes scheduled to be made include the following:

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The peripheral rooms on the north and south sides of the control room complex are being remodeled to improve officiency by rearranging office spaco so that %

functional working groups are near each othe A lunch room, with kitchenette, is being added to the control roo _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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A second lunch room, with kitchenette, is being created in Turbine Hall #3 adjacent to the northwest corner of the control room.

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Two pass-through drawers are being installed in the north wall of the control room com-

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plex to facilitate paperwork transaction Two waiting rooms will be constructed - one outside each pass-through. Closed-circuit -

TV cameras will be installed in each waiting room to permit visual recognitio '

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The control room complex HVAC ductwork con-figuration will be modified to redistribute airflow to the north and south office areas and to better account for equipment heat ;

load The HVAC for the control room pri-mary operating area will be rebalance Smoke detectors will be added, as needed, in the peripheral areas.

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The control room peripheral area lighting is being rearranged to provide a better visual environment. The control room office area lighting is being rearranged to provide indirect lighting for areas with CRis, and direct task lighting and accent lighting where appropriate.

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HVAC, plumbing and electrical services are being provided for the Turbine Hall

lunchroom and the two waiting room The licensee concluded the following items:

MOD 1729 affects the safety-related control

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room HVAC supply system and the control room enclosure walls and floor. The modification i does not involve an unreviewed safety question, t A change to the Technical Specification is required. The modification maintains the capability to safely shutdown the plant in the

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The inspector raised concerns with MOD 1729 regarding control room security changes. This item is addressed in section 10 of this repor . Fuel Bundle 49-40 Removal Fuel bundle #49-40, which was stuck to its fuel support piece, was removed from the core on April 17, 1987, using special procedure (SP)-998, " Removal of Bundle LY6359 in Core Location 49-40", Rev. O. Attempts to dislodge the bundle with a hydraulic jack were unsuccessful. The fuel bundle, with the fuel support piece attached, was placed in a special cask in the fuel pool. GE and the licensee are investigating the stuck bundl The inspector reviewed procedure SF-998, and discussed its implementation with licensed operators and licensee engineers prior to bundle removal from the reactor cor The inspector also monitored portions of the removal and attempts to dislodge the bundle and fuel support piece from the control room and from the fuel floor. The root causes for the bundle and fuel support piece being stuck will be reviewed in a future inspectio No violations were note .4.4 Dropped Cont-ol Rod Bl_ade At 3:08 a.m., on April 22, 1987, a new control rod blade became ungrappled and dropped in the Unit 2 spent fuel pool. The control rod blade was being transported from its storage location in the pool to core location 30-2 Unit 2 was defueled at the time. The control rod blade was being moved with the frame mounted hoist attached to the control rod grapple / latch tool. The control rod blade became lodged at a 20 degree angle with the horizontal, wedged between the south fuel pool wall and the fuel rack identification number plate. No spent fuel nor new fuel is in the area where the control rod blade was resting. No releases of radioactivity occurred and radiation levels in the spent fuel pool area were normal. The licensee investigated the occurrence and developed a procedure to retrieve the control rod blade. The blade was retrieved on April 23, 1987, and was inspected. No apparent damage was noted to the spent fuel pool or the fuel storage racks. The licensee determined that the cause of the ungrappling was apparently a failed dowel pin on the air actuator resulting

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in the failure of the control rod grapple / latch tool to adequately latch the rod bail handle with the "J" hoo The inspector reviewed FH-GE-2, " Replacement and Relocation of Control Rod Blades", Rev. 3; FH-6C, " Fuel Movement and Core Alteration Procedure Ouring a Fuel Handling Outage", Rev. 17; and the Core Component Trans-fer Authorization Sheet (CCTAS). The inspector verified that no radioactive releases occurred by reviewing Unit 2 control room chart traces for the reactor butiding vent stack, reactor zone vent exhaust and refueling floor vent exhaust. The inspector surveyed the dropped control rod blade from the refueling floor. In addition, the inspector discussed the event with licensee engineers, operators, and maintenance personne The inspector questioned whether or not the control rod tool was inspected or tested prior I to use. A licensee representative stated that maintenance procedure M-4.100, " Pre-Outage Preparation Control", Re O, requires inspections and tests of each refueling tool prior to the refueling outage. The inspector reviewed M-4.100 and verified that the control rod tool was tested prior to usage on control rod blades. The inspector also l reviewed the licensee's preliminary investigation. The final investigation and associated report will be reviewed in a future inspectio No violations were note . Remote Shutdown Capability 5.1 Remote Shutdown Panels 10 CFR 50, Appendix A, General Design Criteria No. 19 delineates the Control Room requirement The ability to perform equipment manipulations outside the Control Room to achieve hot and/or cold shutdown are also stated as follows: " Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the t

unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures".

The Peach Dottom Unit 2 and 3 Remoto Shutdown Panels are located at 165 foot level of the Radwaste Butiding. Equipment that can be operated include the following:

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-- RCIC system i

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Safety relief valves E. H L l

._ _____ ____ ______- _-____ _ ___ ___- _

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1 18

<

W

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{ ESW pumps

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Control rod drive hydraulic pumps

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4KV breakers

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Shutdown cooling valves M0-17 & 18.

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4 The Peach Gottom updated FSAR section 7.18 delineates the design

requirements of the Remote Shutdown Control Panels. The FSAR states that the systems to be controlled allow the reactor (s) to i be controlled in a hot standby condition; however, cold shutdown condition cannot be achieved from the remoto shutdown control panel The system is designed only to maintain operations at those remoto panels for one hour, after which Control Room operations are assumed i to be resumed. Technical Specification section 3.11.A, 4.11.A, and

! associated bases state that the Remoto Shutdown Panels are provided 3 to assure hot shutdown condition external to the Control Room. The

! licenseo implements this remoto shutdown capability with procedure

SE-1, " Plant Shutdown from Emergency Shutdown Pano1", Rev 9, dated January 17, 1983.

j lt appears that the as designed and licensed Peach Bottom Remoto Shutdown System does not meet the current 10 CFR 50, Appendix A, General Design Criteria No. 19. The inspector discussed this item

with Ilconsee engincors, operators, and management. The inspector

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reviewed the above stated documentation. This item will be reviewed j further wtth NRR. (!FI 277/87-09-02; 278/87-09-02)

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! 5.2 A_1_tornato Shutdown Method j The inspector reviewed the modifications associated with a1tornato

shutdown system required to comply with the fire protection l requirements of 10 CFR 50, Appendix R. The Peach Gottom Fire

Protection Program Plan describes the alternate shutdown method I

(" Method 0"). A fire in the control room (or cable spreading room i or remoto shutdown panel area) requires the use of the alternate i shutdown method. The method utilizes local control stations i (soparate from the remote shutdown panels on 165 foot level of the radwaste building) as follows:

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-- HPCI alternate control stations (llPCI; RHR; iiPSW; SRVs A, j 0, K; and miscellaneous instrumentation)

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emergency switchgear alternato control stations

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diesel generator alternato control stations j --

ADS alternato control station ;

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I The alternate shutdown method uses HPCI to maintain reactor water level; uses the SRVs to lower reactor pressure; uses RHR and HPSW to remove heat from the torus; and, uses the SRVs, RHR and HPSW for alternate shutdown cooling method. The alternate shutdown cooling method lowers reactor pressure initially to about 50 psig Cold shutdown is achieved by continuing this alternate shutdown cooling method by maintaining the reactor flooded with a RHR pump and removing heat by draining water through the opened SRVs to the toru The SRVs are designed to remain open with the RHR pump discharge pressure providing the necessary backpressure. The RHR pump and HPSW pump are used to remove the heat from the torus (aligned for RHR torus cooling mode).

The inspector discussed this alternate shutdown method and alternate shutdown cooling method with licensee engineers and operators. The inspector also reviewed the Fire Protection Program Report. The alternate shutdown capability will not be operable until modification completion on both units. As stated in licensee letter dated January 8, 1987, the modifications are currently scheduled for completion during the 1987 refueling outages for both Unit No violations were identifie . ReviewofLicenseeEvent__Repolts_(LER_s) LER Review The inspector reviewed LERs submitted tc the NRC to verify that the details were clearly reported, including the accuracy of the description and corrective action adequacy. The inspector determined whether further information was required, whether generic implica-tions were indicated, and whether the event warranted on-site followup. The following LER's were reviewed:

LER N LER Date Event Date Subject 2-85-05, Rev. 1 1RM instrument out of service March 9, 1987 June 12, 1985

• 2-86-06, Rev. 1 PCl3 Group !!! outboard isolation when RPS March 9, 1987 tripped on large motor start February 1, 1986

• 2-86-07, Rev. 1 PCIS Group !!! outboard isolation when RPS March 9, 1987 MG set tripped February 3, 1986

• 2-86-08. Rev. 1 PCl3 Group !!! outboard isolation when RPS MG March 9, 1987 set tripped February 27, 1986

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• 3-86-01, Rev, 1 PCIS Group !!! outboard isolation when RPS March 9, 1987 tripped on large motor start February 3, 1986 2-87-01 Exceeding allowable suppression pool level April 10, 1987 March 11, 1987 2-87-02 Allowable containment leak rato exceeded April 9, 1987 March 13, 1987 6.2 LER On-Site Followup For LERs selected for on-site followup and review (denoted by asterisks above), the inspector vertfled that appropriate corrective action was taken or responsibility assigned and that continued operation of the factitty was conducted in accordance with Technical Specifications and did not constitute an unreviewed safety question as defined in 10 CFR 50.59. Report accuracy, compliance with current reporting requirements and applicability to other site systems and components were also reviewe . LERs 2-86-06, 07, 08 and 3-86-01 were all revised to update Itcensee corrective actions and to further clarify the primary containment isolations. All events occurred when RPS power was tripped either by a loss of RPS MG set (when on normal power) or by a large motor start (when on alternato power). A loss of one RPS power source causes a half scram (no actuations), a half pCIS Group 1 Isolation (no actuations) and a half PCIS Group !!! isolatio The half PCl3 Group !!! isolation results in an outboard or inboard actuation (closure) of secondary containment dampers; and an outboard or inboard actuation (opening) of standby gas treatment system dampers and fan starts, the inspector reviewed each event, and discussed them with Itconseo engineers and operators. The inspector reviewed the corrective actions, procedure changes, and proposed modtfications for the attornate RPS power supply. No inadequacies woro noted relative to these revised LER . Surf S 1_anie Testing The inspector observed surveillance tests to verify that testing had been properly scheduled, approved by shift supervision, control room operators were knowledgeable regarding testing in progress, approved procedures were botng used, redundant systems or components were avall-able for service as required, test instrumentation was calibrated, work was performod by qualifted personnel, and test acceptance critoria were mot. Parts of the following surveillances were observed:

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ST 9.32-2 & 3, Reactor Cold Shutdown Data Log, Rev. 1, performed i

hourly on both Unit 2 and 3 during the period April 7 to April 24, 198 ST 9.12, Reactor Vessel Temperatures, Rev. 8, performed on Unit 3 on April 13, 198 ; ST 9.12C, Reactor Vessel Head Flange Temperatures Surveillance, Rev. 1, performed on Unit 3 on April 13, 198 '

No unacceptable conditions were note . Maintenance i For the following maintenance activities the inspector

spot-checked administrative controls, reviewed documentation, and j observed portions of the actual maintenance

1 Maintenance i

Procedure /

Documen_t E_qu3 ment Dates _ Observed M4.208 Unit 2 RpV head detensioning March 16, 1987 MOD 1958 Remote shutdown panel enhance- April 21, 1987 ments

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! MOD 1790 Unit 2 feedwater heater 1A April 5 and 22, replacement 1987

None Unit 2 high pressure turbine April 15 and 16,

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j inspection 1987  ;

Administrative controls checked included maintenance request forms (MRFs), blocking permits, fire watches and ignition source

, controls, item handling reports, QC involvement, plant conditions, TS LCOs, equipment turnover information, and post maintenance

testing. Documents reviewed included maintenance procedures,

] material certifications, RWPs, MRFs, and receipt inspection '

No inadequactes were Identifie .2 The inspector received information from another DWR regarding sticky lubricant in the main steam isolation valve (MSIV) solenoid valve operators. Apparentlythevalveswerepreviouslylubricated

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with " Parker Super '0" Lube', however the lubricant was changed to i "C. F. Houghton SAr[ 640" due to EQ concerns. This new lubricant i

apparently degrades the aluminum in the AVC0 MS!V actuator mantfold, i l causing friction between the cylinder and piston. The friction could result in restricting movement of the control valve and thus

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] cause slower than desfgn MSIV closure time The inspector reviewed

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maintenance procedure M-1.5, "MSIV Pneumatic Control Manifold Main-tenance", Rev. 9. Procedure M-1.5 requires the use of the " Parker Super "0" Lube" type lubricant which is acceptable based on the i licensee's EQ documentation. The inspector also discussed this item with licensee maintenance personnel. No unacceptable conditions were note . Radiation Protection During the report period, the inspector examined work in progress in both units, including health physics (HP) procedures and controls,

! dosimetry and badging, protective clothing use, adherence to radiation

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work permit (RWP) requirements, radiation surveys, radiation protection instruments use, and handling of potentially contaminated equipment and materials, i

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The inspector observed individuals frisking in accordance with HP procedures. A sampling of high radiation doors was verified to be locked as required. Compliance with RWP requirements was verified during each tour. RWP line entries were reviewed to verify that persunnel had provided the required information and people working in RWP areas were observed to be meeting the applicable requirements. No unacceptable conditions were identifie The inspector reviewed the radiation exposure report dated April 21, 1987. Items checked include exposures to date for the year and quarte No excessive exposures were noted.

10. PJysical Security 10.1 Routine Observations

The inspector monitored security activities for compitance with the accepted Security Plan and associated implementing procedures, including

operations of the CAS and SAS, checks of vehicles on-site to verify proper control, observation of protected area access control and badging procedures on each shift, inspection of physical barriers, checks on control of vital area access and escort procedure No inadequacies were identifie .2 Control Room Modifications The Itcensee is planning to perform modifications to control room por MOD #172 The security related modifications include changes in the north vital area wall by installing pass-through drawer The pass-through drawers will be bullet-resisting in compliance with 10 CFR 73,55(c)(6). Bullet-resistance will be afforded with

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only one (i.e., either the inner or outer) door closed. Closed-circuit TV cameras in the waiting rooms permit visual recognitio The cameras are placed to afford an unobstructed view of the visitor's face and hands. Two way audio communications will be provided. Operation of both inner and outer doors is controlled only

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from the control room side. When the outer door is closed, it locks automatically. An automatic closing mechanism is provided for the outer door. Therefore, when not in use, the pass-through will be closed and locked. During the installation period, a security guard will be posted at the openin ~

The inspector and a region based security specialist raised questions regarding these security changes. Specifically, that the MOD 1729 package did not have any documented security department signoffs nor references to any potential security plan changes. At the end of the inspection, the PORC has not approved MOD 1729 nor the associated safety evaluatio In addition, although plant management and PORC were aware of the proposed changes, the site security organization was unaware of the proposed modification. NRC Inspection 277/87-14 and 278/87-12 also address this issue. This item is unresolved pending further NRC revie (UNR 277/87-09-03; 278/87-09-03)

11. Unresolved Items Unresolved items are items about which more information is required to ascertain whether they are acceptable violations or deviation Unresolved items are discussed in sections 4.4.1, 5.1 and 1 . Management Meetings 12.1 Preliminary Inspection Findings A verbal summary of preliminary findings was provided to the Manager, Peach Bottom Station at the conclusion of the inspectio During the inspection, licensee manag: ment was periodically notified verbally of the preliminary findings by the resident inspectors. No written inspection material was provided to the licensee during the inspection. No proprietary information is included in this repor .2 Attendance at Management Meetings Conducted by Region Based Inspectors Inspection Reporting Date Subject Report No. Inspector 4/20-23/87 Security 87-14/12 Bailey 4/20-24/87 LLRT 87-12/14 Chung 4/20-24/87 HP 87-13/13 Dragoun

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12.3 Attendance At Other Meetings Date Subject 4/10/87 NRC Commissioner briefing and public meeting in Washington, D.C., regarding NRC Order 4/14/87 NRC meeting with the state of Maryland in Annapolis, MD, regarding the NRC Order 4/23/87 PEco meeting and briefing for public officials regarding the NRC Order i