IR 05000277/1986016

Insp Repts 50-277/86-16 & 50-278/86-17 on 860816-0926. Violation Noted:Inadequate Planning & Procedures for DHR & Loss of Shutdown Cooling

ML20215M654
Person / Time
Site:	Peach Bottom
Issue date:	10/24/1986
From:	Eselgroth P, Hillman B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20215M646	List: IR 05000277/1986016 ML20215M645
References
50-277-86-16, 50-278-86-17, NUDOCS 8611030133
Download: ML20215M654 (26)

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U. S.' NUCLEAR REGULATORY COMMISSION

REGION I

-Report No. 50-277/16 & 50-278/17 Docket No. 50-277 & 50-278 License No. 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: August 16 --September 26, 1986 Inspectors: T. P. Johnson,-Senior Resident Inspector J. H. Williams, Resident Inspector Reviewed By:  % /Mwu / /d/7//6g B. M. Hillmap', Reactor EngiTneer date Approved By: // f[

P. W. Esp)'groth, Chief 'dat4 DRP, Section 2A

. Inspection Summary: Routine, on-site regular and backshift resident inspection (138 hours0.0016 days <br />0.0383 hours <br />2.281746e-4 weeks <br />5.2509e-5 months <br /> Unit 2; 122 hours0.00141 days <br />0.0339 hours <br />2.017196e-4 weeks <br />4.6421e-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, outage activities, maintenance, emergency drill, and outstanding item Results: Inadequate planning and procedures for decay heat _ removal and loss of shutdown cooling were identified (detail 4.2.1). Unit 3 was manually scrammed due to a resin injection from a condensate demineralizer (detail 4.2.3). Good operator response to a Unit 3 condensate problem (detail 4.2.2) and to a Unit 3 recirculation pump trip (detail 4.2.4) were noted.

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DETAILS Persons Contacted J. B. Cotton, Superintendent Plant Services

• R. S. Fleischmann, Manager, Peach Bottom Atomic Power Station A. A. Fulvio, Technical Engineer A. E. Hilsmeier, Senior Health Physicist J. F. Mitman, Maintenance Engineer D. L. Oltmans, Senior Chemist F. W. Polaski, Outage Planning Engineer S. R. Roberts, Operations Engineer D. C. Smith, Superintendent Operations S. A. Spitko, Administration Engineer J. J. Tucker, Emergency Preparedness Coordinator 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 cold shutdown for performance of environmental qualification (EQ) inspections of Limitorque motor operated valves (MOV). On August 17, 1986, an inadvertent heatup and slight pressurization occurred (see detail 4.2.1). Unit 2 restarted on August 26, 1986, after all EQ inspections were complete The unit continued to operate at or near full power during the remainder of the report period, except for a load drop en September 20, 1986 for performance of Surveillance testing on the MSIVs and reacto .2 Unit 3 Unit 3 also began the inspection period in cold shutdown in the performance of EQ inspections of Limitorque MOVs. Unit 3 startup commenced on August 28, 1986; however, a blown fuse in the hotwell level controller caused an unplanned CST overfill (see detail 4.2.2) that terminated the startup. The unit subsequently restarted on August 29, 1986. Load drops occurred on September 6, 9, and 11, 1986, for control rod pattern adjustments and maintenanc ,- -. . -. .- , -.- .

On September 14, 1986, Unit 3 was manually scrammed due to high main steam line high radiation (see detail 4.2.3). The unit restarted on September 15, 1986; however, the unit was shut down on September 16, 1986, due to a HPCI M0-15 packing leak. Following repairs the unit was restarted on September 17, 198 On September 24, 1986, the 3A reactor recirculation pump tripped due to loss of the 3A stator coolant pump (see detail 4.2.4). The recirculation pump was restarted and the unit was returned to full powe . Previous Inspection Item Update 3.1 (Closed) linresolved Item (277/85-44-01). Failed DC solenoids in Main Steam Isolation Valves (MSIVs). A scram on Unit 2 occurred on January 24, 1986, when two MSIVs closed because power was lost to the AC coil combined with two failed DC coils. Similar DC coil failures resulted in scrams on Unit 3 on July 11, 1984 and July 19, 1986 (see detail 6.2.1). The licensee replaced the failed DC coils and has determined that the failure mechanism appears to be thermal life related as the coils are continuously energized in a

- thermally hot area (135 degrees F). The licensee has revised the environmental q'ualification preventive maintenance (EQ PM) to replace these DC coils on all MSIVs every refueling. The inspector verified the PM frequency by reviewing the EQ equipment lists and through discussions with licensee engineers. In addition, the licensee has revised the periodic monitoring of the MSIV AC and DC coils by implementing procedure ST 21.9, MSIV Coil Continuity Checks. The inspector noted that this ST was performed on both Unit 2 and 3 MSIV coils on August 25, 1986. Based on the review of the LERs for the three scrams, en the EQ PM program for DC coil replacement, on the review of ST 21.9, and on discussions with licensee engineers on failure mechanisms, the item is resolved and therefore close .2 (Closed) Unresolved Item (277/80-17-01). MSIV solenoid valve replacement and documentaticn review. The MSIV pilot solenoid valves were replaced during the 1980 Unit 2 refueling outage. The inspector reviewed modification package #72E and related documentatio No unacceptable conditions were identified, therefore the item is l

closed.

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3.3 (Closed) Unresolved Item (277/80-17-02). MSIV pneumatic control l manifold cracking repair. During the 1980 Unit 2 refueling i

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outage, the licensee repaired the cracking per Procedure M1.5, MSIV Pneumatic Control Manifold Maintenance. The inspector reviewed the completed procedure and MR No unacceptable conditions were identified, therefore this item is closed.

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3.4 (Closed) Unresolved Item (27'//84-15-01). Licensee response to the April 22, 1984 earthquake felt on site. The licensee responded to the earthquake indications by implementing procedure SE-5,

" Earthquake" and by monitoring seismic instrumentation. The licensee reported the earthquake initially to the NRC, and in a follow-up letter dated May 11, 1984. The inspector was concerned with the inoperability of both the active and passive seismic instrumentation, and with the operators unfamiliarity with seismic monitoring system operation. The inspector reviewed the licensee's response and discussed the item with licensee engineers and operator ~

The licensee revised the seismic s; tem operating procedures to include more detail. Additional surveillance procedures were written and implemented to perform periodic checks of the passive seismic monitors (triaxial peak accelerographs). Additional training was performed to ensure operations personnel are familiar with seismic event follow-up procedures and actions. Additional seismic monitoring system reviews are documented in detail 4.5 of this report. Based on the review of the licensee response, on interviews with licensee personnel, and on the seismic system walkdown, the item is resolved and therefore close .5 (Closed) Unresolved Item (278/85-44-04). 10 CFR 21 followup on Bonney Forge pipe fittings used on Unit 3 scram discharge volume instrument volume (SDVIV). A weldolet used in modifications to Unit 3 SDVIV in 1981 is the subject of the unresolved item. The licensee determined that the weldolet was purchased to requirements in excess of the ASME code and that it was not reportable under 10 CFR 21. The inspector reviewed licensee engineering and QA documentation and analysis, and discussed the item with the licensee. No unacceptable conditions were noted, therefore, this item is close .

3.6 (Closed) Inspector Follow Item (277/82-25-05; 278/82-24-02).

Follow long term corrective actions of LER 2-83-1/IP regarding ECCS room cooler switche To ensure adequate cooling of the ECCS rooms (including HPCI, RCIC, core spray, and RHR rooms), the licensee installed jumpers and lifted leads in response to environmental qualification (EQ) and fire protection separation .

criteria concerns in January 1983. The licensee has completed room cooler related modifications on both units as follows: MOD

1. 1174 to change the room cooler control switches to ensure EQ of the switch cams and retainer rings; MOD #10298 to ensure fire protection separation criteria; and, MOD #1364 to replace the room cooler differential pressure switches to meet EQ criteria. The inspector reviewed the jumper and lifted lead log, reviewed the related modification packages, and reviewed system operating and surveillance test procedures. Selected ECCS rooms were inspected to verify adequate positioning of the room cooler switche .

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The inspector also discussed the modifications and cooler system operation with licensee engineers and operators. No unacceptable conditions were identified, therefore, this item is close .7 (Closed) Unresolved Item (277/86-13-01; 278/86-14-01). Drywell Chilled Water and Reactor Building Closed Cooling Water (RBCCW)

Primary Containment Isolation Valves. The licensee reviewed these manual remote isolation valves and concluded that revisions to operating procedures were necessary. The licensee revised GP-8, Primary Containment Isolation and TRIP Procedure T-101, Reactor Pressure Vessel Controls, to include manual closure of these containment isolation valves. The inspector reviewed a licensee mechanical engineering letter and station engineering memo both dated August 22, 1986; and the revised GP-8 and T-101 procedure The inspector has no further concerns, therefore, this item is close .8 (Closed) Unresolved Item (277/86-13-03; 278/86-14-03).

Environmental qualification (EQ) inspections of Limitorque motor operated valves (MOV) internal wiring. The licensee completed the EQ inspections on both units during this inspection period. The inspector reviewed the Unit 2 and Unit 3 summary reports which

- describe the results of the EQ MOV inspections. The inspector also reviewed the Safety Evaluation for MOD 2062, Revision 1, Supplement 1, August 22, 1986. This safety evaluation justifies not performing the EQ inspections on Unit 3 MOVs M0-3-10-17 and 18. These two valves are the RHR shutdown cooling suction valve The licensee decided not to perform the EQ inspections because of reactor water temperature control problems experienced on Unit 2 (see detail 4.2.1). The licensee's justification for not performing EQ inspections on MO-3-10-17 and 18 included: (1) no unqualified wiring had been found on any of the Unit 2 and 3 EQ inspections, (2) the safety function of the valves is to close on PCIS signals and the valves are normally closed, (3) the valves should open to establish shutdown cooling, and (4) if the valves were to fail to open, alternate shutdown cooling could be established per TRIP procedure T-115, Alternate Shutdown Coolin The inspector attended Plant Operations Review Committee (PORC)

meeting #86-113 on August 26, 1986. At this meeting the PORC reviewed the Unit 3 EQ MOV inspection summary reports; approved safety evaluation for MOD 2062, Revision 1, Supplement 1; and, determined that Unit 3 could startup and return to power operation. The inspector verified that the PORC meeting was in accordance with Technical Specifications 6.5.1 and procedure A-4, PORC Procedure.

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Based on review of M00 2062 safety evaluation; review of the EQ MOV inspection results, and PORC meeting #86-113, this unresolved item is closed. The inspector will continue to follow the EQ inspection of the remaining two MOVs (M0-3-10-17 and 18) during the next Unit 3 refueling outag .9 (Closed) Violation (277/84-16-01; 278/84-14-01). Failure to have whole body counting procedure The licensee responded to the violation in a letter dated July 13, 1984. The inspector reviewed the response and determined it to be adequate. The inspector also reviewed the HP0/CO-26 series procedures which include the whole body counting implementing procedures. No unacceptable conditions were noted, therefore, the violation is close .10 (Closed) Violation (277/84-14-02; 278/84-12-02). Inadequate corrective-actions for deficiencies identified during QA audits on equipment storage and open engineering review request forms. The licensee responded to the violation in a letter dated July 18, 1984. The inspector reviewed the licensee response and discussed it with licensee QA personnel. No unacceptable conditions were noted, therefore, the violation is close .11 (Closed) Unresolved Item (277/84-14-01; 278/84-12-01). QA organization in the QA Plan, QA Program Description, and the FSA The 1983 QA Program Description did not reflect the current QA organization. The licensee revised the QA organization in the QA Plans (Volumes I and III), in the QA Program Description, and in the FSAR. The inspector reviewed the current revisions to the QA Plans Volume I and III, the QA Program Description, and the updated FSAR (1986 revision). No unacceptable conditions were

noted, therefore, the unresolved item is close .12 (Closed) Unresolved Item (277/85-04-02). Radiographic indication anomalies transverse to the weld length of 1.5" thick shop longitudinal welds in stainless steel 28" 0.D. pipe. Tne inspector reviewed the General Electric Company memorandum dated February 22, 1985 (Ishizaka to Lebre) indicating the transverse indications to be due to X-Ray diffraction and not an unacceptable metallurgical or weld integrity condition. A sample of pipe weld with this condition was submitted to the Oak Ridge National Laboratory (ORNL) for independent examination and analysis. The results of the ORNL work on the submitted sample confirmed that no adverse conditions existed that would be of concern to the fitness of the pipe for reactor service'. The ORNL report was transmitted to NRC

RI by memorandum dated August 4, 1986, from Serpan to Durr.

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This report and the above referenced G. E. report were reviewed by the inspector. The inspector has no furthte concerns, therefore, this item is close _ , - _ . _ _ _ _ _ -

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3.13 (Closed) Inspector Follow Item (277/84-22-01). . Review licensee's actions to train Health Physics personnel in procedure A-86,

" Administrative procedure for corrective action of idnetified discrepancies". The licensee is developing a procedure " Health Physics Deficiency Report" which will give guidance to health physics personnel on the use of Health Physics deficiency reports. The new procedure was reviewed in NRC Inspection 277/86-18; 278/86-1 This inspector follow item is closed, however, the inspector will follow the implementation of the new Health Physics Deficiency Report system as part of routine inspection .14 (Closed) Unresolved Item (277/81-24-02). ECCS logic power supply failures. Units 2 and 3 experienced ECCS logic power supply (ELMA) failures in 1981 and 1982. The power supply provides redundant 24 Volt DC power to the ECCS initiation instrument There are two power supplies (C722A,B) for each unit, and each power supply has redundant components (E/S-402/3A,B). The licensee performed modification (MOD) #976 in March 1983 to replace the ECCS power supplies for both units. These new G. power supplies have not experienced similar failures. The licensee continues to perform a monthly surveillance test (ST),

ST-1.14 A(B), "ECCS A(B) Power Supply Ripple" which verifies power supply operability by checking output voltage. The inspector reviewed MOD #976 documentation, reviewed ST-1.14A(B), inspected panels C722A(B) and power supplies E/S 402/3A and B, and discussed the item with licensee engineers. Based on the above, this item is close . 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 for compliance with 10 CFR 50.54 and Technical Specifications. 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 that indicated values were within Technical Specification requirements and normal operating limit ECCS switch positioning and valve Ifneups 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

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were knowledgeable of alarm status and plant condition . The inspector checked for fluid leaks by observing sump status, alarms, and pump-out rates; and discussed reactor coolant system leakage with licensee personne The Unit 3 limit of 2.0 gpm unidentified leakage per NRC-Order was verified frequentl . Shift relief and turnover activities were monitored daily, including backshift observations, to ensure compliance tith administrative procedures and regulatory

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i guidanc No inadequacies were identified.

l 4. The inspector observed the main stack and both reactor l building ventilation stack radiation monitors and

[ recorders, and periodically reviewed traces from backshift periods to verify that radioactive gas release rates were within limits and that no unplanned releases l had not occurred. No inadequacies were identified.

l l 4. The inspector observed control room indications of fire i 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 observed fire-fighting equipment station No inadequacies were identified (see detail

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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 after maintenance. Plant housekeeping improvements continue, and overall plant appearance is goo . The inspector observed the nuclear instrumentation subsystems (source range, intermediate range and power range monitors) and the reactor protection system to verify that the required TS channels were operabl .1.10 The inspector frequently verified that the required TS off-site electrical power startup sources and emergency on-site diesel generators (DG) were operable. On September 23, 1986, the E-3 DG was taken out of service for the annual preventive maintenance. The inspector verified that TS 3.5.F.1 (seven day LCO) and TS 4.5. (daily DG and ECCS testing) were followed. The inspector noted that the requirement to perform daily ECCS testing requires starting 24 safety related pumps (four core spray, four RHR and four high pressure service water for each unit). This requirement is not consistent with standard technical specifications which do not require the daily testing of ECCS pumps. The inspector discussed this potential unnecessary testing with NRR and the licensee, and will review this in a future inspection (IFI 277/86-16-02). No violations were note .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 l drawings, instrumentation and breaker positioning.

l Selected major components were visually inspected for l

leakage, proper lubrication, cooling water supply,

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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 Unit 2 and 3 RCIC systems. No inadequacies were identifie .

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4.2 Followup On Events Occurring During the Inspection 4. Unit 2 Heatup To Greater Than 212 Degrees F and Pressurization While In Cold Shutdown Condition 4.2. Summary On August 17, 1986, the Unit 2 reactor coolant water temperature was inadvertently allowed to rise to 245 degrees F while the unit was in cold shutdown without primary containmen Several Technical Specification limiting conditions for operation (LCO) action statements were entered because of the heatup above 212 degrees F. The reactor coolant temperature remained above 212 degrees F for a period of not greater than 11 hour1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> .2. Sequence of_ Events On the morning of August 17, 1986, the unit

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was in the cold shutdown mode for inspections and rewiring of Limitorque valve operators to ensure compliance with equipment environmental qualification (EQ) requirements. The reactor mode switch was in shutdown with the reactor pressure vessel head vent valves A0-17 and A0-18 open to vent the reactor to the drywell equipment drain sum The condensate system was on long path recirculatio The reactor water cleanup, control rod drive, and reactor recirculation system pumps were not operating because of the EQ inspections. The A and B

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loops of core spray and the A loop of RHR-LPCI l were out of service. The "D" RHR pump was

! operating in the shutdown cooling mode with high pressure service water flow varied through the "D" RHR heat exchanger for reactor coolant temperature control. Reactor coolant temperatures were measured at the "D" RHR pump I

discharge. This was the only operable coolant l temperature measurement system because the recirc pumps were secured. Primary containment I

integrity was not required below 212 degrees F and not established. The standby gas treatment system was operating, taking suction from both the Unit 2 and Unit 3 drywell Secondary con-tainment was not required but established.

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The licensee planned to remove RHR shutdown cooling from service for about six hours to allow inspection and rewiring of the RHR shutdown cooling suction valves M0-17 and M0-18. These valves would be blocked closed for the EQ inspections. General procedure GP-12, Core Cooling Procedure, Rev. 6, dated April 21, 1986, would be used for decay heat removal while the RHR shutdown cooling mode was out of service. Reactor water level would be raised to the main steam lines and drained to the main condenser. Operation of the condensate system would allow make up and reactor coolant water level contro Reactor coolant temperature would be inferred from reactor pressure vessel skin temperatures (thermocouples). It was expected by the licensee that raising the water level above

+55 inches would allow natural circulation flow and adequate mixing to prevent stratification. There were no alternative or

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backup methods to cool the reactor by approved normal operating procedures. However, TRIP procedure T-115, Alternate Shutdown Cooling, was availabl At 9:45 a.m., the reactor operator began to slowly increase reactor water level being careful to keep the reactor pressure vessel (RPV) head flange temperature above 100 degrees F. At 12:35 p.m., the water level was at +55 inches and coolant temperature was 165 degrees F at the "0" RHR heat exchanger inle At this time, the "D" RHR pump was removed ,

from service to allow blocking the shutdown cooling suction valves M0-17 and M0-18. ST 9.12, Rev. 8, " Reactor Vessel Temperatures",

which records reactor pressure was stopped as allowed at 12:45 p.m., because shutdown cooling was secured and reactor water level was greater than +50 inches. The reactor vessel head temperature increa ed from 154 degrees F to 190 degrees F between 1:00 and 2:00 p.m., as recorded on ST 9.12c, Rev. O, " Reactor Vessel Head Flange Temperature Surveillance". Since reactor vessel level at

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this time was about 140 inches below the flange, the temperature increase was likely to be caused by stea .- - .- .

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Reactor water level was +100 inches at 3:00 Forty-five minutes later the reactor operator noticed the drywall equipment drain sump was automatically pumping out. Input to the sump was via the open manual valve (HV-19)

in the two inch line between the "C" main steam line and the reactor vessel head vent line which is piped to the sump. The reactor coolant water level at this time was up to the main steam lines (110 inches). Between 3:45 p.m. and 6:10 p.m., the drywell sump automatically pumped out several times. At 6:10 p.m., an operator was sent into the drywell to close HV-19 because of the frequency of the sump pump out Reactor vessel water level was 120 inches at 7:00 Within the next half hour the reactor oper.ator noticed reactor pressure was 17 psig. Control room personnel concluded that reactor coolant had stratified and bulk

- coolant temperature was greater than 212 degrees F. To measure coolant temperature and promote mixing within the reactor vessel the licensee decided to start the A recirculation pump. At 9:05 p.m., the A recirculation pump was started and reactor coolant temperature was measured as 230 degrees F at the pump suctio The Shift Superintendent conferred by phone with the Operations Engineer at 9:15 p.m. It was decided to drain reactor water to the torus and make up with cooler water from the condensate system. Two " letdowns" were completed using the "B" RdR loop through the torus full flow test line. Charts from the torus level reactor LR 8027 indicate these l letdowns occurred-at approximately 9:30 and 10:30 p.m. At 10:50 p.m., PORC approved special procedure SP-947, Alternate Shutdown Cooling Using Condensate Feed with Vessel Drain to Torus Via RHR. At 9:30 p.m., the i

recirculation pump suction temperature was 245 degrees After the two letdowns to the torus, the reactor water temperature was 220 degrees F at 11:00 p.m. Torus level was at the Technical Specification limit of 14.9 feet before the letdowns and subsequently increased to 15.25 fee . _ ._

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At about 11:00 p.m. M0-18 and M0-17 were returned to service. RHR shutdown cooling was established at 11:40 p.m., and by 11:45 p.m.,

the reactor coolant water temperature was less than 212 degrees TS LCOs for primary containment, torus water level, low pressure ECCS were thus satisfie .2. NRC Review and Conclusions The inspector reviewed the licensee's investigation and reports, including the draft Upset Report, Event Report No. 23, and LER 2-86-20 associated with this event. The inspector reviewed documents listed in Attachment 2 and discussed the event with plant operattng and engineering personne The licensee concluded that Unit 2 was operated without adequate planning and with inadequate procedures for decay heat removal and shutdown cooling. Technical Specification - 6.8.1, ANSI N18.7-1972, and Regulatory Guide 1.33 (November 1972) require nuclear power

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plants to be operated in accordance with written procedures. These procedures shall be sufficiently detailed for a qualified individual to perfor Failure to have adequate procedures for decay heat removal and shutdown cooling mode of operation is a violation of TS 6. No Notice of Violation is being issued because: (1) the licensee identified the occurrence; (2) the event fits Severity Level IV or V; (3) LER 2-86-20 has been submitted and reviewed; (4) corrective actions have occurred or are planned; and (5) corrective actions for a previous violation in 1981 (LER 2-81-31) would not have prevented this violatio In addition, all TS LCO action statements were met and the licensee's investigation and reporting of the event was complete. The operations group prepared a draft Upset Report, the ISEG group prepared Event Report No. 23, and LER 2-86-20 was submitted. Licensee corrective actions include the following:

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A new procedure will be written to address the removal of RHR shutdown cooling from servic GP-12 will be revised to provide additional guidance for decay heat remova ST 9.12 will be revised to require reactor temperature and pressure monitoring when reactor level is greater than 50 inche Discussion of the event with plant operators, engineers, and supervisio Include the event in operator training programs.'

The inspector will review these corrective actions in a future inspection (IFI 277/86-16-01).

4.2.2 Unit 3 Condensate Storage Tank (CST) Overfill on August 28, 1986 The Unit 3 reactor operator had just begun to withdraw control rods for reactor startup on August 28, 1986, when he received an alarm indicating low condenser hotwell level "C". The level appeared satisfactory so he tried to reset the alarm, but it would not clea Shortly thereafter the "A" and "B" hotwell low level alarms were also received. Operators determined within minutes that the fuse in the hotwell recorder circuit in i panel 30C07A had blown. The fuse was replaced and blew again. The fuse is in the circuit of ten other I recorders and controllers for panel 30C07A. The condensate pump was tripped at 6:53 a.m., to stop the flow to the CST. The operator began to equalize the level between the refueling water storage tank (450,000

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gallons) and the CST (200,000 gallons) and took l appropriate corrective action to maintain reactor vessel l coolant level. The blown fuse caused the hotwell makeup

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the condensate pump discharge header to the CST to open.

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The sequence is similar to an event in February 1986 that resulted in a spill to the river (see NRC Inspection 278/86-05). Personnel were dispatched to the Unit 3 CST to determine if a spill had occurred and sample any water on the ground within the dike

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er J-structur Sample results indicated that only rain water was presen Licensee estimates indicated approximately 2000 gallons of storage capacity was left before the CST would have spilled through the top ven Overflow to the radwaste collector tank was 11,115 gallons. A short was found in the ribbon cable for the hotwell level recorder LR-3085 and the cable was replaced. MOD 1926, " CST Fused Power Feeders for Miscellaneous Recorders and Controllers" was performed prior to restarting the uni MOD 1926 provides additional instrument power wiring and separate fuses for each instrument and controller inside control panel 30C07 The inspector was in the control room and witnessed operator response to this event. Operators took timely and appropriate action during the event. Their response was excellent. The inspector discussed the event with the operators and reviewed documentation for MOD 1926, including MRF 305C8601664 for the work. The inspector also examined the new circuits inside control panel 30C07A. The licensee has scheduled M00-1926 for Unit 2 during the next refueling outage. No unacceptable conditions were note .2.3 Unit 3 Manual Scram on September 14, 1986 Unit 3 was manually scrammed by the reactor operator at 5:10 p.m., on September 14, 1986, from 57% power due to main steam line (MSL) high radiation. The MSL high radiation was caused when a resin injection occurred when the 3H condensate demineralizer was placed in service. Plant operators initially reduced power from 100% by reducing recirculation flow and inserting cor, trol rods when the MSL radiation monitors increased to the alarn setpoint (1.5 times normal). When the "C" MSL radiation monitor spiked (3.0 times normal) a half reactor auto scram and half group I (MSIV) primary containment isolation occurred. The operators manually inserted a reactor scram by placing the reactor mode switch to shutdown per procedure OT-103, "MSL Radiation". The MSL radiation monitors increased to about 2200 mr/hr (3.0 times normal is 2500 mr/hr).

Reactor conductivity increased to 13.0 mhos/cm and pH decreased to 4.6 due to the resin injection. Group II and III primary containment isolations occurred after the scram when reactor water level decreased below four inches. The reactor feed pumps subsequently recovered reactor water level. The licensee made an ENS call at 6:25 p.m., on September 14, 1986, and notified the Senior Resident Inspector at about 6:00 p.m. A small

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release (normal on the scram) occurred out the main stack which was about 2% of the Technical Specification limi The plant remained in hot shutdown at about 270 degrees Two RWCU pumps and demineralizers reduced the conductivity to less than 1.0 mho/c The licensee determined the cause of resin injection to be unseating of several elements in the 3H condensate filter demineralizer uni On September 15, 1986, the inspector reviewed control room indications, instrument chart traces, sequence of events, computer log and operator logs; and, discussed the event with the operators. Operator response was in accordance with emergency procedures OT-103, T-100, and T-99. The licensee prepared an Upset Report and completed GP-18, Scram Review Procedure. The inspector reviewed these documents and discussed them with the licensee. The inspector discussed the chemistry excursion with licensee engineer Within the scope of the review of this event, no

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violations were note .2.4 Unit 3 3A Recirculation Pump Trip On September 24, 1986 At 11:05 p.m., on September 24, 1986, with Unit 3 at 100% power, the 3A reactor recirculation pump trippe The apparent cause of the trip was loss of the 3A main generator stator coolant pump cn motor fault. The 3B stator coolant pump started in standby; however, the pressure surge in the stator coclant system resulted in a 3A recirculatioa pump trip. The recirculation pumps are designed to trip on time delay after a loss of the stator coolant syste Reactor power decreased to 70% upon the re:frculation pump 3A trip. The plant operators followed procedures OT-112, Recirculation Pump Trip, Rev. 3 and OT-113, Loss of Stator Cooling, Rev. O. Reactor power was further reduced to less than 35% and the tripped 3A recirculation pump was restarted in accorcance with S.2.3.1.A, Startup of a Recirculatten Pump, Rev. 11, at 11:48 p.m., on September 24, 1986. The unit was then returned to full powe The inspector reviewed procedures OT-112, OT-113, and S.2.3.1.A and discussed their implementation with the licensed operators. The inspector also reviewed control room chart recorders and traces, control room logs, computer sequence of events log; and, discussed the

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event with operations personnel. The inspector concluded that the operator response was good and timely; and, in accordance with emergency and system operating procedure Within the review of the 3A reactor recirculation pump trip event, no unacceptable conditions were identifie .3 Logs and Records The inspector reviewed logs and records for accuracy, completeness, 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 and STA Log Book, Night Orders, Radiation Work Permits, Locked Valve Log, Maintenance Request Forms, Temporary Circuit Modification 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. No unacceptable conditions were identifie .4 Unit Startups Following EQ Inspections Unit 2 started up on August 26, 1986, and Unit 3 started up on August 29, 1986, upon completion of the EQ f r.spections of Limitorque motor operated valve internal wiring (see detail 3.8).

The inspector observed portions of both Unit startups including:

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Unit 2 heat up and pressurization,

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Unit 2 power ascension,

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Unit 3 initial rod withdrawal and criticality,

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Unit 3 heat up and pressurization,

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Unit 3 power ascensio The inspector reviewed the general plant startup procedures including GP-2 series and individual system startup procedure Electrical Production Department QC was noted as observing both unit startups. The rod worth minimizer system was operable and functioning adequately during the startups. The inspector

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verified that a second licensed operator was present during the startup as required by Technical Specifications. Adequate supervision including Shift Superintendent, Shift Supervisor, and licensee management was present during each startup as verified by the inspector. No unacceptable conditions were note .5 Seismic Monitoring Instrumentation The inspector performed a detailed walkdown of portions of the seismic monitoring system in order to independently verify the operability of the Unit 2 and 3 common system. The walkdown included verification of the following items:

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Review of documents listed in Attachment Inspection of system equipment conditions; including passive and active detectors, and seismic panel 00C75 Confirmation that the system check-off-list (COL) and operating procedures are consistent. with plant drawing Verification that system breakers and switches are properly aligne Verification that instrumentation is operabl Verification that control room and local switches, indications and controls are satisfactor Verification that surveillance test procedures properly implement the Technical Specifications surveillance requirement ,

No violations were identifie . Emergency Plan Drill on September 18, 1986 The licensee cenducted an emergency plan drill on September 18, 198 The purpose of the drill was as follows:

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To provide training in preparation for the October 1986 annual emergency exercise.

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To meet the periodic drill requirements of Confirmatory Action Letter (CAL) #85-17 dated November 5,198 To provide training for the minimum shift manning requirements of NUREG-0654 as stated in PECo letter dated August 8, 198 _

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To verify adequacy of the new position of shift technical assistant to perform dose assessment and calculation To verify the adequacy of the shift technical advisor (STA) to function as the core physics / thermal hydraulic adviso The inspector attended the pre-drill briefing on September 17, 1986, and observed portions of drill on September 18, 1986. Observations included the following:

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Control room response, assessment and mannin Control roo.i initial dose assessment by the shift technical assistan Fire drill resoonse by the fire brigad Technical Support Center and Emergency Operations Facility response, assessment and mannin Through observations and interviews, the inspector concluded that the STA can adequately perform the functions required within the first hour of an accident declaration required by NUREG-0654 in the area of core physics / thermal hydraulics. However, the inspector noted that the STA duties and responsibilities as stated in procedure A-7, Shift Operations, do not specifically state the core physics / thermal hydraulics function. The inspector discussed this with the licensee, and licensee engineers indicated that procedure A-7 would be reviewed and revised accordingly. The inspector will review the revised procedure A- Within the scope of the review of the emergency drill, no violations were note . Review of Licensee Event Reports (LERs)

6.1 LER Review The inspector reviewed LERs submitted to 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 implications were indicated, and whether the event warranted on-site followup. The following LER's were reviewed:

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LER N LER Date Event Date Subject 2-86-17 Control rod block monitor system August 4, 1986 July 4, 1986 2-86-19 Exceeding leakage results for type B/C September 5, 1986 local leak rate tests August 8, 1986

• 2-86-20 Heatup above 212 degrees F during cold September 22, 1986 shutdown August 17, 1986

• 2-86-21 Unauthorized work on core spray minimum September 12, 1986 flow valves August 13, 1986 ,

• 3-86-16 Unit 3 scram when #3 startup source was lost August 18, 1986 and one MSIV close July 19, 1986

• 3-86-17 Unit 3 hourly fire watch not performed September 5, 1986 August 7, 1986

• 3-86-18 Unit 3 scram on low level when 3C RFP tripped September 10, 1986 August 11, 1986 6.2 LER On-Site Followup

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For LERs selected for on-site followup and review (denoted by asterisks above), the inspector verified that appropriate corrective action was taken or responsibility assigned and that continued operations of the facility 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 . LER 2-86-21 concerns unauthorized work on the Unit 2 core spray minimum flow valves during the environmental qualification inspections. The event was reviewed in NRC Inspection 277/86-13; 278/86-14. There were no inadequacies relative to this LE P

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6. LER 3-86-16 concerns a Unit 3 scram when the #3 off-site power source was lost combined with the closure of the 80B MSIV due to a failed DC coil. The event was reviewed in NRC Inspection 277/86-13; 278/86-14. The DC coil failure is discussed in detail 3.1 of this repor There were no inadequacies relative to this LE . LER 2-86-20 concerns a heat up on Unit 2 above 212 degrees F during cold shutdown (see detail 4.2.1).

6. LER 3-86-18 concerns a Unit 3 scram on low level when the 3C RFP tripped on thrust bearing wear detector. The event was reviewed in NRC Inspection 277/86-13; 278/86-1 There were no inadequacies relative to this LE . LER 3-86-17 concerns Technical Specification fire watch patrols which were not performed for approximately 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> on August 7 and 8, 1986. The inspector discussed the event and the corr.ective actions with the license The fire patrol instructions for Units 2 and 3 were reviewed. Technical Specification paragraph 3.14. . requires the licensee to verify the operability of appropriate fire detectors when fire barriers are not functional. The inspector questioned the licensee's method of verifying fire detector operabilit The licensee checks to determine that fire detectors are in the area of the degraded fire barrier and that the surveillance test for the detector is curren After discussions with regional specialists, it was con-cluded that the licensee's actions were adequate. The inspectors had no further questions or concerns with this even . Surveillance Testing The inspector observed surveillance tests to verify that testing had been properly scheduled, approved by shift supervision, control room

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operators were knowledgeable regarding testing in progress, approved

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procedures were being used, redundant systems or components were available for service as required, test instrumentation was calibrated,

work was performed by qualified personnel, and test acceptance criteria were met. Parts of the following tests were observed

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ST 6.11, RCIC Pump, Valve, Flow & Cooler, Revision 30, 5/9/86, performed on Unit 3 on September 16, 198 ST 2.5.28, Functional & Instrument Check of the Seismic Monitoring System, Revision 5, 6/30/86, performed on Unit 2 on September 22, 1986.

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In addition, a review of selected completed surveillance tests listed in Attachments 1 and 2 was performe No inadequacies were identifie . Maintenance For the following maintenance activities the inspector spot-checked administrative controls, reviewed documentation, and observed portions of the actual maintenance:

Maintenance Procedure /

Document Equipment Date Observed M3. Hydraulic contrql unit level September 11, 1986 (MRF N switch replacement for control 3M8606851) rod #58-31 Administrative controls checked included maintenance request forms (MRFs), blocking permits, fire watches and ignition source controls, item handling reports, QC involyement, plant conditions, TS LCOs, and equipment turnover information, and post maintenance testin Documents reviewed included maintenance procedures, material certifications, RWPs, MRFs, and receipt inspection No inadequacies were identifie . 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 work permit (RWP) requirements, radiation surveys, radiation protection instruments use, and handling of potentially contaminated equipment and material 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 personnel had provided the required information and people working in RWP areas were observed to be meeting the applicable requirements. No unacceptable conditions were identifie . Physical Security The inspector monitored security activities for compliance 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

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and badging procedures on each shift, inspection of physical barriers, checks on control of vital area access and escort procedures. No inadequacies were identifie . In-Office Review of Public and Special Reports The inspector reviewed the 1985 Annual Plant Modification Report for Units 2 and 3. The report is required by 10 CFR 50.59 and includes:

engineering and research modifications, station modifications, and tests / experiments. No unacceptable conditions nor inaccuracies were note . Inspector Follow Items Inspector follow items are items for which the current inspection findings are acceptable, but due to on going licensee work or special inspector interest in an area, are specifically noted for future follow-u Follow-up is at the discretion of the inspector and regional managemen Inspector follow items are discussed in details 4.1 10 and 4. . Management Meetings 13.1 Preliminary Inspection Findings A verbal summary of preliminary findings was provided to the Manager, Peach Bottom Station and to the Superintendent, Nuclear Generation Division, at the conclusion of the inspection. During the inspection, licensee management was periodically notified verbally of the preliminary findings by the resident inspector 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

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Inspection Reporting Date Subject Report N Inspector l

Aug 25 - Security 277/86-17 Bailey

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28, 1986 278/86-18 Sep 8 - Health Physics 277/86-18 Dragoun 12, 1986 278/86-19 i

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ATTACHMENT 1 Seismic Monitoring Documentation

-- FSAR Appendix C, Structural Design Criteria

-- FSAR Section 2.5, Geolooy and Seismology

-- FSAR Section 7.21, Suismic Instrumentation

-- Technical Specification 3.15/4.15, Seismic Monitoring Instrumentation

-- SE-5, Earthquake, Revision 4, 10/7/85

-- NRC Regulatory Guide 1.12, Instrumentation for Earthquakes

-- Abstract 13.6, Peak Recording Accelerographs and the Triaxial Strong Motion Seismic Monitoring System, Revision 1

-- S 13.6, Operation of the Seismic Monitoring Panel 00C755, Revision 1, 5/17/85

-- S 13.6-1, Analysis of Seismic Peak Accelerographs After An Event, Revision 0, 1/21/82

-- ST 2.5.31, Calibration of the Triaxial Peak Accelerographs (PAR 400),

Revision 1, 1/3/86

-- ST 2.5.33, Functional Test of Response Spectrum Analyzer (RSA-50),

Revision 2, 12/16/85

-- ST 2.5.34, To Calibrate Response Spectrum Analyzer (RSA-50), Revision 1, 12/16/85 -

-- ST 2.5.36, Instrument Check of PAR-400, Revision 2, 5/22/86

-- ST 2.5.35, Seismic System Trigger Card Calibration, Revision 2, 12/16/85

-- ST 2.5.28, Functional & Instrument Check of the Seismic Monitoring System (Active), Revision 5, 6/30/86

-- Technical Specification Amendment Nos. 75 and 74 for DPR-44 and DPR-56, dated 11/19/80

-- E-111 Drawings, Wiring Diagram Seismic Monitoring Instrumentation

-- Modification Package #609

-- Alarm Cards 30C212L #4 and #5

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ATTACHMENT 2 Documents Reviewed During Followup To Unit 2 Heatup On August 13, 1986 Upset Report (Draft), Rev. 7, dated 8/22/86, " Sequence of Events Leading to Unit 2 Reactor Exceeding 212 Degrees Fahrenheit on 8/17/86" Peach Bottom Event Report #23, " Unit 2 Reactor Water Temperature Rise Resulting in Technical Specification Violations" I&E Circular 81-11, July 24, 1981, " Inadequate Decay Heat Removal During Reactor Shutdown" I&E Information Notice 86-74, August 20, 1986, " Reduction of Reactor Coolant Inventory Because of Misalignment of RHR Valves" GE SIL #251, October 31, 1977, " Control of RPV Bottom Head Temperatures GE SIL #251, Supplement 1, July 1980, "BWR Vessel Bottom Head Coolant Temperature Measurement ( AID 46-79)"

GE SIL #357, June 1981, " Control of Reactor Vessel Temperature / Pressure During Shutdown" GE SIL #401, December 1983, "BWR Shutdown Cooling - AID 68" NSAC 27, September 1981, " Analysis of Heatup and Pressurization During Dresden-3 Shutdown" NSAC 88, March 1986, " Residual Heat Removal Experience Review and Safety Analysis - Boiling Water Reactors" INPO SOER 82-2, April 28, 1982, "Iaadvertent Reactor Pressure Vessel Pressurization" Memorandum from S. Roberts, 8/18/86, " Exceeding 212 Degrees Fahrenheit On Unit 2 On August 17, 1986" Franz-0-Gram, August 15, 1986 LER 2-81-031-3L-0, June 17,1981, " Coolant temperature exceeding 212 degrees F without primary containment" NRC Inspection Report 277/81-14 and 278/81-15, July 29, 1981 PECo letter dated 8/27/81, from Daltroff to Keimig NRC Inspection Report 277/83-37 and 278/83-35, February 16, 1984

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Operating Logs - Unit 2 Operators Log, Control Room Shift Supervisor's Log, and Unit 2 Drywell Health Physics Log MRF #2-10-C8606048 for M0-17 MRF #2-10-C8606254 for M0-18 General Procedure, GP-12, Rev. 6, dated 4/21/86, " Core Cooling Procedure" Special Procedure SP 947, Rev. O, August 21, 1986, " Alternate Shutdown Cooling Using Condensate Feed with Vessel Drain To Torus Via RHR" RT 3.0, Rev. 6, February 3,1983, " Plant Water Inventory" ST 9.12, Rev. 8, April 21,1986, " Reactor Vessel Temperatures", performed 8/11/86 to 8/17/86 ST 9.128, Rev. 3, April 29, 1986, " Reactor Coolant Temperaturs", performed 8:25 p.m. and 9:00 p.m., on 8/17/86 ST 9.12C, Rev. O, September 1,1983, " Reactor Vessel Head Flange Temperature Surveillance", performed 8/11/86 to 8/18/86 TR 2-02-3-89, Temperature recorder data on reactor vessel skin temperature for 8/17/86 Chart from LR 8027 Unit 2 torus level - narrow range for period of 8/17/86 Chart from TR/LR-81238 Unit 2 torus temperature and level for period of 8/17/86 Drawing M-295, sheet 78, Rev. 19, " Critical Small Piping - Isometrics Inside Drywell #2" Drawing M1-R-1, Rev. 12, " Primary Steam Piping" Drawing M-351, Rev. 24 " Nuclear Boiler" Drawing M-361, Rev. 28, " Residual Heat Removal System" Drawing M1-B-65, Rev. 14, " Reactor Assembly"