IR 05000282/1988012

Insp Repts 50-282/88-12 & 50-306/88-12 on 880625-0806. Violations Noted.Major Areas Inspected:Previous Insp Findings,Plant Operational Safety,Maint,Surveillances,Esf Sys,Ler Followup & Preparation for Refueling

ML20153D358
Person / Time
Site:	Prairie Island
Issue date:	08/22/1988
From:	Burgess B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20153D351	List: IR 05000282/1988012 ML20153D349 ML20153D354
References
REF-GTECI-124, REF-GTECI-A-26, REF-GTECI-NI, REF-GTECI-RV, TASK-124, TASK-A-26, TASK-OR 50-282-88-12, 50-306-88-12, GL-87-06, GL-87-6, IEB-87-002, IEB-87-2, IEIN-87-004, IEIN-87-4, NUDOCS 8809020178
Download: ML20153D358 (13)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Reports No. 50-28?/88012(DRP); 50-306/88012(ORP)

Docket Nos. 50-232; 50-306 Licenses No. DPR-42; OPR-60 Licensee: Northern States Power Company 414 Nicollet Mall Minneapolis, MN 55401 Facility Name: Prairie Island Nuclear Generating Plant Inspection At: Prairie Island Site, Red Wing, Minnesota Inspection Conducted: June 26 through August 6, 19tB Inspectors: J. E. Hard M. M. Moser Approved By: . Burgess 2 8Y Reactor Projects Section 2A Date Inspection Summary Inspection on June 25 through August 6, 1988 (Reports No. 50-282/88012(DRP);

50-306/88012(ORP))

Areas Inspected: Routine unannounced inspection by resident inspectors of previous inspection findings, plant operational safety, maintenance, surveillances, ESF systems, LER followup, preparation for refuelinc. Information Notice and Bulletin followup, closecut of Temporary Instructions, and Gene"ic Letter followu Results: During this inspection perind, both units operated continuously at 100% power except for Unit 1 conuencing coastdown and also experiencing a 39 hour4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br /> forced outage to modify Foxboro Style "C" controllers. In general the plant continues to be operated well as noted by no reactor trips since July 1987, a decline in personoel errors, good equipment condition being maintained, and excellent progress being made in preparing for the upcoming Unit 1 outag However, certain deficient conditions identified in this report appear to have existed for a very long tim Specific details can be found in Paragraphs 3 and 1 This raises some concerns about the adequacy and thoroughness of the licensee's efforts in such areas as surveillance tests, design bases reviews, and in ensuring that all 10 CFR and technical specification requirements are in fact being fully addresse Of the ten areas inspected, two violations of NRC requirements were identified; one in the area of plant operational safety (Paragraph 3) involving the inoperability of three of four RPS channels due to a generic design deficiency in Foxboro Style "C" controllers, and the other involved review of unresolved safety issue A-26 (Reactor Vessel Pressure Transient Protection For Pressurized Water Reactors; Paragraph 10) in which the pressurizer power operated relief valves (PORV) were not included as part of ASME Section XI code requirements and consequently had not been stroke time tested until very recentl In addition, an unresolved item was identified in the area of surveillances (Paragraph 5) involvi.g the adequacy of response time testing that was being performed and may also be a generic issu G809020178 880322 PDR ADOCK 03000282 '

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DETAILS Persons Contacted

• E. watzl, Plant Manager

• 0. Mendele, General Superintendent, Engineering and Radiation Protection R. Lindsey, Assistant to the Plant Manager

• H. Selle,an, General Superintendent, Operations D. Schuelke, Superintendent, Radiation Protection G. Lenertz, General Superintendent, Maintenance

• K. Beadell, Superintendent, Technical Engineering

• M. Klee, Superintendent, Quality Engineering R. Conklin, Supervisor, Security and Services D. Vincent, Project Manager, Nuclear Engineering and Construction '

J. Goldsmith, Superintendent, Nuclear Technical Services A. Hunstad, Staff Engineer .

T. Amundson, Superintendent Training

• A. Smith, General Superintendent, Planning and Services

• E. Eckholt, Senior Nuclear Safety / Technical Services Engineer A. Vukmir, Site Services Representative, Westinghouse Electric Cor F. Gavigan, Westinghouse Electric Cor O. Dilanni, License Project Manager, NRR The inspectors interviewed other licensee employees, including members of the technical and engineering staffs, shift supervisors, reactor and auxiliary operators, QA personnel, Shift Techaical Advisors, and Shift Manager * Denotes those present at the exit interview of August 8,198 . Licensee Action On Previous Inspection Findings (92701)

(Closed) 282/87014-01(DRP) Open Item: Lube Oil Problems With No. 11 Auxiliary Feedwater Pump On September 30, 1987, during routine surveillance testing of No. 11 turbine-driven auxiliary feedwater pump, the auxiliary oil pump failed  !

to stop after the feedwater pump came up to speed. Investigation revealed <

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that air binding of the gear pump was the cause and that this was due to excessive running of the auxiliary oil pump. The running of the auxiliary

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oil pump is controlled by a timer and is intended to ensure that the

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auxiliary feedwater pump thrust bearing remains adequately wetted with l lube oil. After discussions with the pump manufacturer and additional

testing it was found that shorter auxiliary oil pump runs eliminated air binding of the gear pump. Replacement of the auxiliary oil pump timers on all four auxiliary feedwater pumps will be completed prior i

! to the Unit i refueling outage in August 198 J (Closed) 282/87016-06; 306/87015-01(DRP) Violation: Failure of the Licensee to follow Their Own Procedures Regarding Escorting Visitors

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During November 1987, region based examiners identified a discrepancy in the licensee's escort practices while administering examinations to senior reactor operator candidates at the plant. Specifically, it was ,

found that escorts were permitting visitors to badge into vital areas i.

' first in lieu of the correct procedure which was to have the escor; hadge in firs This resulted in a Level V violation. The licensee has rev1>cd Procedure SAWI5.1.1 (Revision 1) to permit visitors to badge in before or after their escor (Closed) 282/87016-05(ORP) Violation: Failure to Follow Written Procedure When Cutting and Removing Electrical Cable On October 19, 1987, craft personnel were removing abandoned electrical cabling in the auxiliary building and cut the wrong cable while making an intermediate cut. Cause of this error was a failure to follow written procedures and resulted in a Level IV violation. Corrective action was promptly taken to repair the incorrectly cut cable, revise the work request to include signoffs, and retrain all individuals involve In addition, this incident received wide dissemination to all craft and supervisory personnel to reinforce the importance and safety significance of this even (Closed) 282/88004-03; 306/88004-02(ORP) Open Item: Trending of Check Valve Leak Rates Licensee demonstrated that trend tables for selected Section III valves were current and being maintained up to dat (Closed) 282/87012-04; 306/87011-01(ORP) Open Item: Possible Leaky Seals in Core Exit Thermocouple (CET) System Elastomer seals in the CET connectors were not fully qualified for the containment post-accident environment. This lack of qualification required the assumption of significant errors in the thermocouple signals. All unqualified seals in both units were replaced with properly qualified one (Interim Report) 282/85024-04; 306/85022-04(ORP) Open Item: Post-Accident Emergency Cooling Water Flow Requireinent and Availability In a meeting with the resident inspectors on August 9, 1988, the licensee committed to provide by September 17 a report summarizing and analyzing their position on questions raised by the inspectors regarding cooling water system adequac The questions being reviewed related to: Adequacy of a single diesel-driven cooling water pump to provide the cooling *1ater loads during loss-of-offsite power situations simultaneous with various plant condition Capacity of the emergency cooling water line with one diesel-driven cooling water pump as compared to the flow required to be availabl N

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. The extent of river silting, its effect on river water available to the emergency cooling water intake, and corrective action planne . Operational Safety Verification (71707, 93702)

Unit 1 was base loaded at 100% power except for reductions for surveillance testing until July 10, 1988, when Unit 1 commenced coastdown for a refueling outage. Unit 1 also experienced a 39 hour4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br /> forced outage on July 26, 1988 while a field change was made to Foxboro Style C entrollers. Unit 2 was base loaded at 100% power except for reduction for surveillance testin The inspector observed control room operations, reviewed applicable logs, conducted discussions with control room operators, and observed shift turnover The inspector verified operability of selected emergency systems, reviewed equipment control records, and verified the proper return to service of affected components. Tours of the auxiliary building, turbine building and external areas of the plant were conducted to observe plant equipment conditions, including potential fire hazards, and to verify that maintenance work requests had been initiated for equipment in need of maintenanc On July 1, 1988 with both units at 100% power, the licensee found an open penetration below motor control center (MCC) No. INA2 located on 755 level in the auxiliary building whicn penetrated the auxiliary building special ventilation zone (A85VZ). The open penetration measured approximately three square feet. Further investigation revealed one additional NCC with an unsealed base. It was found that neither open penetration violated fire protection zones and their combined open areas did not exceed the allowable ten square feet permitted for the ABSV However, to improve performance of the ABSVZ test, both MCC open penetrations were seale On July 25, 1988, Unit 1 was at 85% power (end-of-cycle coastdown) and Unit 2 was at 100% power. At 1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br />, a load reduction to 50% power was begun on Unit 1 for maintenance. The time spent at reduced power was longer than anticipated, and as xenon concentration was increasing, it became necessary to withdraw control rods to maintain reactor coolant system temperature and power. As a result of the control rod withdrawal and the xenon build-up, a large axial flux tilt developed which ultimately reached +24%.

The flux tilt was sufficient to require the reactor protection system to apply a penalty to the overpower and overtemperature delta T reactor protection setpoint. However, operators noted that one of the four channels (the blue channel) of overpower and overtemperature delta T setpoint was not responding properly. Investigation overnight and the l following day revealed that the blue channel flux tilt controller did not respond to changes in the input signal and that the affected flux tilt controller was a Foxboro 62-H-2E Style C, while the other three Unit 2 controllers were Foxboro 62-H-2E Style The Style B controllers responded properly. When the corresponding controllers in Unit 2 were i

inspected, three of the Style C controllers, which would not respond properly to flux difference input and one Style B controller, which responded properly, were found in us .

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At 2324 hours0.0269 days <br />0.646 hours <br />0.00384 weeks <br />8.84282e-4 months <br /> on July 26, a Notification of Unusual Event (NUE)

was declared and a shutdown of Unit 2 was begun since the minimum operable delta T reactor protection channel requirements of Technical Specification 3.5 were not me Three of the four channels of each delta T trip function are required to be operabl Unit 1 was then shutdown so that the Style B controllers installed in Unit 1 could be removed and transferred to Unit 2 (which was capable of more generating capacity since Unit 1 was in end-of-cycle coastdown). Unit I was in hot shutdown by 2400 hours0.0278 days <br />0.667 hours <br />0.00397 weeks <br />9.132e-4 months <br />. By 0146 hours0.00169 days <br />0.0406 hours <br />2.414021e-4 weeks <br />5.5553e-5 months <br /> on July 27, the Unit 2 nuclear instrumentation system (NIS) delta T channels again satisfied the Technical Specification requirements for minimum operable channels. The NUE and Unit 2 load decrease were terminated and Unit 2 was returned to full powe Following consultation with Westinghouse and Fox'.]oro, a modification to the Style C controllers was developed by the plant technical staff and reviewed by the Operations Committe The four Style C controllers were modified to make them compatible with the Prairie Island nuclear instrumentation system (supplied by Westinghouse). The modified Style C controllers were installed in Unit 1 and tested. Unit I was returned to service at 1340 hours0.0155 days <br />0.372 hours <br />0.00222 weeks <br />5.0987e-4 months <br /> on July 28. Preliminary investigation indicates that Unit 2 appears to have had three of four channels inoperable since 1974 and that Unit 1 may have had two of four channels inoperable during routine channel surveillance testing and maintenance. This is a violation of Technical Specification 3.5 which requires three of four channels to be operable. See Notice of Violation (306/88012-01(ORP)). Maintenance Observation (62703)

Routine, preventive, and corrective maintenance activities were observed /

reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides, and industry codes or standards, and in conformance with Technical Specifications. The following items were considered during this review: the limiting conditions for operation were met while components or systems were removed from service, approvals were obtained prior to initiating the work, activities were accomplished using approved procedures and were inspected as applicable, functional testing and/or calibrations were performed prior to returning components or systems to service, quality control records were maintained, activities were accomplished by qualified personnel, radiological controls were implemented, and fire prevention controls were implemented.

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l Portions of the following maintenance activities were observed / reviewed

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during the inspection period:

l Heater Drain Pump repair Control Room carpet replacement

' Foxboro Style C trouble shooting and testing Unit 2 pressurizer spray control valve testing Emergency Diesel No. 4 preop testing Maintenance of security fence l

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Update of Auxiliary Feedwater System Commitments:

In a letter dated November 26, 1986, the NRC detailed its assessment of the overall reliability of the auxiliary feedwater (AFW) systems at Prairie Island relating to Generic Issue No. 124, "Auxiliary Feedwater System Reliability." The review concluded that the AFW systems are adequately designed, maintained, and operated and adequately address the concerns of Generic Issue 124 provided that a list of eight recommendations noted in the report are implemented by the license A review by the resident inspector has determined that the licensee has incorporated all of the NRC recommendations and this is to be documented in the next annual revision to the updated safety analysis report (USAR).

The results of a probabilistic risk assessment (PRA) initiated by the licensee for the auxiliary feedwater system (AFW) in April 1986 generated a list of 19 candidate modifications and recommendations which would improve AFW system reliability. All but two of the 19 items will have been incorporated in both plants by the end of the Unit 1 refueling outage in September 198 In addition to the NRC and PRA generated recommendations noted above, the licensee is also incorporating a number of upgrades and improvements in an ongoing program to further improve the AFW system reliabilit Substation Proble,ms:

8H1781 Motor-Operated Disconnect - This disconnect for one of the Unit 1 main Generator 345 kv output breakers failed to operate when called on to do s Investigation revealed that the thermostat-heater arrangement intended to keep moisture out of the motor enclosure had misoperated causing the motor to overheat thus preventing operation. Wiring insulation was melted inside the box. Thermostat failure along with wrong size heater seem to be the specific causes. Licensee reports that there probably has been no PM program for these component No violations or deviations were identifie . Surveillance (61726)

The inspector witnessed portions of surveillance testing of safety-related systems and components. The inspection included verifying that the tests were scheduled and performed within Technical Specification requirements, by observing that procedures were being followed by qualified operators, that Limiting Conditions for Operation (LCOs) were not violated, that system and equipment restoration was completed, and that test results were acceptable to test and Technical Specification requirement Portions of the following surveillances were observed / reviewed during the inspection period:

• SP 1747 Fuel Oil Storage Tank Leak Test

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• SP 1002A Unit 1 Analog Protection Test

• SP 1116 Unit 1 Flux Mapping

• SP 11058 No. 12 Diesel Cooling Water Pump Test

• SP 1728 Siren Cancel Test Response Time Testing Introduction The inspectors followed up on Northern States Power Company's (NSP) Nuclear Operations Quality Assurance (N0QA) Audit AG 87-43-3 entitled "Response Time Testing" (RTT) that was intended to determine

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if the instrument RTT performed at the NSP nuclear plants were being performed in accordance with the nuclear plant's Technical Specifications (TS). Testina The current Prairie Island TS's requires RTT of the nuclear power range, intermediate range, source range, reactor coolant overtemperature Delta T, overpower Delta T, and the reactor trip breakers. Testing of the above instrumentation was performed in-accordance with Table TS.4.1-1 and TS Definition E.4, "Channel Response Test." The Definition states, "A channel response test consists of injecting a simulated signal into the channel as near the sensor as practicable to measure the time for electronics and relay actions, including the output scram relay."

The inspectors reviewed Surveillance Procedure SP 1008 (2008),

"Reactor Protection Logic Time Response Test," to ensure the licensee was testing the required instrument channels per Table TS 4.1-1 and TS Definition E.4 and this was found to be the case, Transient Analysis The Basis for channel response tests was described in TS 4.1-3 which states, "Measurement of response times for protective channels are performed to assure response times within those assumed for accident analysis (FSAR, Section 14)." The inspectors reviewed NSP's calculation methods and input parameters that were used for the Prairie Island safety and accident analysis as stated in USAR Chapter 14. The USAR states, in part, "the DYN0DE model for the Prairie Island Plant is based on the Technical Specification limits and the assumptions used are consistent with those used in the FSAR analysis." The associated time delays for each trip function used in the DYNODE model was listed in USAR Table 14.3- The DYNODE-P program is a digital computer program that models the behavior of pressurized water reactors under normal and abnormal operating condition [, l

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The following is a list of input parameters and their delay time

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assumed in the safety analysis:

TRIP FUNCTION DELAY TIME (SEC.) SCHEDULED TESTING High Neutron Flux Yes Low Reactor Coolant Flow No High Pressurizer Pressure No low Pressurizer Pressure No High Pressurizer Water Level No low-Low Steam Generator Water Level No Overtemperature Delta T Yes Overpower Delta T Yes The above trip functions that are currently not tested were tested according to TS Definition E.4 in Preoperational Test 25.3, "Reactor Protection Time Response," in August of 1973. However, the licensee !

has assumed manufacturer's specified response times for the process sensors (pressure and Delta P transmitters) that sense pressure, level, and flo This time was assumed to remain constant, was not verified during preoperational testing, and the total response time (delay time) for these instrument channels are currently not tested to support the safety analysi The licensee is currently reviewing actual plant data to try and determine actual sensor and channel response times. This effort is being conducted by the NSP Nuclear Analysis Department (NAD)

for both the Monticello and Prairie Island sites. The inspectors requested the licensee to provide the following information for both sites:

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(1) Determine the process instrument channels that were assumed to mitigate the most limiting transients analyzed in the USA (2) Determine the total channel response times for these instrument I channels from plant data, type testing data, or actual performanc l (3) Determine from the above data, that the plants are not !'

operating in an unanalyzed conditio l

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(4) Determine if changes in response time testing methodology are l warranted.

, (5) Determine if any changes are required to the safety analysis,

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Technical Specification response time definition, Technical i

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Specification Basis, and surveillance testin l (6) Develop a schedule for implementing any of the above change l l 8 l

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The above items are considered a part of Unresolved Item 282/88012-01 and 306/88012-02. The licensee's NAD committed to the NRC to complete the analysis by October 31, 1988, for both Prairie Island and Monticello sites. The NRR Licensing Project Managers for Prairie Island and Monticello concurred with the commitment date and with Region III followup on the licensee's actions at that tim This question of response time testing may be a generic one which could apply to many of the older plants, Summary

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, inspector's review of the process sensor instrument records adicated the transmitters were being calibrated (per Technical

$pecifications), were receiving adequate maintenance, and exhibited stable operation. The licensee's N0QA audit was found to include both a programmatic and technical review. The technical review was in-depth and the audit recommendations were based on goo.1 engineering judgemen No violations or deviations were identifie . ESF System Walkdown (71710)

The inspector performed a complete walkdown of the accessible portions of Unit 1 and Unit 2 Emergency Diesel Generators. Observations included confirmation of selected portions of the licensee's procedures, checklists, plant drawings, verification of correct valve and power supply breaker positions to insure that plant equipment and instrumentation are properly aligned, and local system indication to insure proper operation within prescribed limit No violations or deviations were identifie . Preparation for Refueling (60705)

Preparations for the upcoming Unit I refueling and maintenance outage commencing August 23 are underwa The inspectors observed surveillance testing of fuel receipt handling equipment and new fuel receipt inspectio Unit I will receive 48 new Westinghouse fuel assemblies during this outage which with the reactor core reshuffle will yield 440 effective fuel power (EFP) days. The refueling outage is scheduled to take 34 days and in addition to refueling will include:

Steam Generator Eddy Current Inspection and Sleeving Removal of LIO Reactor Head Cap Accumulator Nozzle Inspection No. 12 Reactor Coolant Pump Motor Retainer Ring Replacement Steam Generator Crevice Flushing No. 12 Auxiliary Feedwater Pump Lube Oil Cooler Modification NSSS Annunciator Panel Modifications Add Seal Weld Over Instrument Column Canopy Integrated Leak Rate Test of Containment

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Motor Operated Valve Actuator Testing Control Room "E" Panel Modifications Component Cooling Heat Exchanger Repairs Secondary Side Pipe Thinning Inspection No. 12 Circulation Water Pump Impeller Replacement Inservice Testing 8. Licensee Event Reports Followup (92700)

Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine that reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with Technical Specifications:

(Closed) LER 282/87008-LL: Autostart of No. 12 Diesel Cooling Water Pump On June 8, 1987, a routine operability test of No. 22 diesel cooling water pump was underway. Near the end of the test the No. 21 motor driven cooling water pump was started to assume the load of No. 22 diesel cooling water pump prior to it being shut down. When the speed of the No. 22 pump was being reduced, the No. 12 diesel c.coling water pump auto started due to low pressure in the cooling water header. Autostart of No. 12 diesel cooling water pump was caused by air binding of No. 21 motor driven pum Long term corrective action involved revising the surveillance procedures (i.e. SP 11061b, Rev. 20) and normal operating procedures (i.e. C35, Rev. 5) to provide additional instructions / guidance to the operators concerning possible air binding in the motor driven cooling water pump (Closed) LER 282/87011-LL: No. 2 Emergency Diesel Generator Thrust Bearing Replacement During the annual maintenance inspection of No. 2 emergency diesel generator in March 1987, the lower main thrust bearing (No.13) was found to be damaged. Further investigation found the damage was due to a breakdown of the generator insulation allowing stray current arcing to damage the bearing. Following repairs, No. 2 diesel was successfully run-in and declared operabl Tests subsequently performed on the No. 1 emergency diesel generator found the generator insulation to be in good condition. The licensee has revised maintenance procedures to include insulation checks during future annual inspection (Interim Report) LER 282/87015-LL: Severe Weather Caused Partial Loss of Offsite Power On July 27, 1987, a tornado caused power to be interrupted on one of the two 345 kv distribution line This resulted in lockout of No. 10 transformer, one of the Murces of offsite power to the emergency buse _ _ - . _-

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a The operable diesel generator started and the inoperable one returned to service while the event was in progress. Much electrical equipment

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tripped during the event and was restored manually. One cooling water pump tripped, causing automatic start of No.12 diesel driven cooling water pump, because the screen house bay level transmitter wa's not provided with emergency powe Many other lessons were learned which are not discussed in this LER and the licensee has agreed to provide a revisio . Information Motice Followup (92701, 25593)

(Closed) NRC Information Notice (IN) No. 87-04: Diesel Generator Fails Test Because of Degraded fuel (TI 2515/93)

A review of the licensee's QA program for emergency diesel generator fuel oil by the resident inspector has found that: Diesel generator fuel oil ic obtained from Q-listed source New deliveries are sampled per Procedure C-38 and analyzed to verify the correct fuel oil was receive Periodic sampling of all diesel fuel oil storage and day tanks is performed to verify that the fuel oil is in good condition and meets Technical Specification 4.6. The inspector concludes that the licensee has an adequate emergency diesel fuel oil QA program in plac It should be noted that as part of the effort to add two additional emergency diesel generators (D-5, D-6)

at this plant, the licensee will be conducting an extensive review with contractor assistance of diesel fuel storage facilities and safeguards necessary to prevent degraded fue (Closed) 282/87002-18; 306/87002-1B; NRC Bulletin 87-02, Supplements 1 and 2:

Fastener Testing to Determine Conformance with Applicable Material Specifications As noted in Inspection Reports No. 50-282/88004(ORP); and No. 50-306/88004(DRP), the licensee issued a detailed report on January 26, 1988 of the information that was requested in NRC Compliance Bulletin 87-02. On April 22, 1988, NRC Bulletin 87-02, Supplement I was issued requesting the licensee submit additional information on the source of fasteners purchased for use in nuclear power plant On June 10, 1988, NRC Bulletin 87-02, Supplement 2 was issued clarifying the type of information requested in Supplement On July 13, 1988, the licensee responded to NRC Bulletin 87-02, Supplements 1 and 2 by providing a list of fastener suppliers and subvendors for safety related and non-safety related procuremen _ _ _ - - - _ _ _ _ _ _ _ _ _ - - - - ,

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10. Closeout of Temporary Instruction (TI) (25019)

Unresolved Safety Issue A-26: Reactor Vessel Pressure Transient Protection For Pressurized Water Reactors, TI 2500/19 Inspections required by this TI were completed with the following results:

05.01 Design Both units at Prairie Island received their operating licenses prior to March 14, 1978 and consequently are permitted to use a combination of operator manual and automatic equipment responses to protect the reactor vessel from low terrperature overpressurization events. A review of the design change that covers the pressure protection system (Design Change 77L416)

and related correspondence found that the system meets the intent of the design requirements. This includes features to compensate for a single failure in addition to a failure that initiated the pressure transient as well as features to compensate for a pressure transient and a failure of equipment needed to terminate the transien .02 Administrative Controls ynd Procedures Plant operations procedures were reviewed and found to contain adequate instructions and precautions with respect to RCS low-temperature overpressure concerns during reactor startup, operation, shutdown as well as emergency operatio .03 Training and Equipment Modifications Operators receive classroom instruction and simulator training for RCS low-temperature overpressure protection system operatioh as well as receiving information from event reports during refresher training.

l, l 05.04 Surveillance A review of the surveillance procedures associated with the overpressure protection system found them to be generally acceptable except for surveillance procedure SP 2291 (pressurizer power operated r111ef valve stroke timing - Unit 2).

This surveillance is a very new one and appears not to have been performed on either unit prior to January, 1988. Investigation has found that the pressurizer power operated relief valves (PORVs) had not been considered part of ASME Section XI and that, therefore stroke time testing was not required. This is contrary to Technical Specification 4.2.A which requires inservice inspection of ASME Code Class 1, Class 2, and Class 3 components in accordance with Section XI. See Notice of Violation (282/88012-02; 306/88012-03(DRP)).

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11. Generic Letter Followup (92701)

(Closed) 282/87006-HH; 306/87006-HH, Generic Letter 87-06: Periodic Verification of Leak Tight Integrity of Pressure Isolation Valves This generic letter requested information regarding all pressure

~ isolation valves in the plant. This information was provided to NRC within the 90 days specified in the lette . Exit (30703)

The inspectors met with the licensee representatives denoted in Paragraph 1 at the conclusion of the inspection on August 8, 1988. The inspectors discussed the purpose and scope of the inspection and the findings. The inspectors also discussed the likely information content of the inspection report with regard to documents or processes reviewed by the inspector during the inspection. The licensee did not identify any document /

processes as proprietary.

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