IR 05000271/1987006

Insp Rept 50-271/87-06 on 870303-0406.No Violations Noted. Major Areas Inspected:Actions on Previous Insp Findings, Physical Security,Plant Operations & Preparation for Shipment of Irradiated Matls from Spent Fuel Pool

ML20209D833
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	04/22/1987
From:	Elsasser T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20209D823	List: IR 05000271/1987006 ML20209D820
References
50-271-87-06, 50-271-87-6, NUDOCS 8704290372
Download: ML20209D833 (17)

Text

. . - . .-= . ... .. . . . - -- -

.

.

U. S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N Docket N License No. DPR-28 Licensee

Vermont Yankee Nuclear Power Corporation RD 5, Box 169, Ferry Road Brattleboro, Vermont 05301 Facility: Vermont Yankee Nuclear Power Station Location: Vernon, Vermont j Dates: March 3 - pril 6, 1987

Inspector: William . Raymond Se or Resident Inspector

'

Approved by: / '

Thomas C. Elsasser, thykt, Reactor Projects Section 3C Date l Inspection Summary: Inspection on March 3 - April 6, 1987 (Report No. 50-271/87-06)

r Areas Inspected: Routine, unannounced inspection on day time and backshifts by the resident inspector of: actions on previous inspection findings; physical security; plant operations; preparation for shipment of irradiated materials from the spent fuel pool; surveillance testing; maintenance activities; proposed installation of a high density storage rack as a design change to the spent fuel storage pool; personnel and proposed organization changes; containment requirements for RHR pump wear ring inspections; followup of audit findings; and, operator attentiveness to

licensed dutie The inspection involved 142 hours0.00164 days <br />0.0394 hours <br />2.347884e-4 weeks <br />5.4031e-5 months <br />.

' Results: No violations were identified in 11 areas inspected. Routine reviews of plant activities identified no conditions adverse to safe plant operations.

i i

!

!

i

,

l l

8704290372 870423 1 i

'

i PDR ADOCK 05000 G

f

.

-- .- _ _ _ . -- . - . - -_ . . . - . - _ - - - . - _ -

.

.

DETAILS Persons Contacted Interviews and discussions were conducted with members of the licensee staff and management during the inspection period to obtain information pertinent to the areas inspected. Inspection findings were discussed periodically with the management and supervisory personnel listed belo Mr. R. Lopriore, Maintenance Supervisor Mr. R. Pagodin, Technical Services Superintendent Mr. J. Pelletier, Plant Manager Mr. J. Sinclair, Security Supervisor Mr. R. Wanczyk, Operations Superintendent Mr. T. Watson, Instrument and Control Supervisor The inspector contacted the Vermont State Nuclear Engineer by telephone on several occasions during the inspection period to discuss inspection findings and areas of plant operations of mutual interest, and to review proposed NRC inspection plans of licensed activities at the plan . Summary of Facility Activities The plant continued routine operations at rated power during the inspection period. A meeting with licensee management was held on March 27, 1987 to discuss the results of the NRC staff's Systematic Assessment of Licensee Per-formance for the period of October 19, 1985 to December 31, 1986. The turbine generator was removed from the electrical grid during a controlled shutdown on April 4,1987 to repair a weld leak in the 1 inch above-seat drain line for the #2 turbine control valv .0 Status of Previous Inspection Findings 3.1 (0 pen) Unresolved Item 86-25-02: Bingham RHR and Core Spray Pump Part 21 Reportability. The licensee informed the inspector on March 18, 1987 that the results of his engineering evaluation of the Bingham-Willamette minimum flow requirements have concluded that the item is reportable

, under 10 CFR Part 2 NRC Region I inspection 86-25 summarizes the in-i

'

itial notification from the licensee regarding the potential for dele-terious pump effects due to inlet / outlet recirculation flow phenomena for minimum flows less than 38% of design. The licensee's evaluation concluded that no safety concern exists at Vermont Yankee because the Core Spray and RHR pumps, which have minimum flow recirculation lines set for 5% of design flow, are not operated in the minimum flow mod The licensee concluded that the matter should be reported under Part 21 ,

since a significant hazard may exist at another facility where the '

Bingham pumps are used in a configuration that provides extended opera- l tion at low flow '

i

---

. . . - - , _ - .

~ 1

-t The inspector reviewed the Part 21 written report submitted on March 20, 1987 and verified that it accurately summarized the event. During a conversation with the Senior Engineer (0perations) on March 25, 1987, the licensee highlighted the following typographical error in the text:

the reference to "2075" in the second line from the bottom of the first page of Enclosure 1 to the March 20th letter should read "29,200" hours allowable. The inspector noted that the typographical error was readily discernible since 2075 is previously defined in the text as the pump minimum flow in gallons per minute for intermittent pump operation. No further licensee action is required to correct the Part 21 repor The item remains open pending completion of licensee actions to address other concerns raised in Inspection 85-25 and subsequent review by the NR '

3.2 (Closed) Unresolved Item 87-04-01: Main Steam Line Alarm Setpoint Change This item is addressed further in Section 12.0 below. The inspector determined that completion of the setpoint change initiated as a result of the surveillance test on February 17, 1987 was not necessary due to a change in the full power background reading on the "C" main steam line

,

radiation monitor. The inspector noted further that the monitor high

,

' alarm and high-high trip setpoints were proper for the background read-ings observed on March 12 and 13, 1987. This item is close .3 (0 pen) Violation 86-25-01: Failure to Meet ASME Section XI Requirements for Inservice Testin The licensee responded to this item by letter FVY 87-30 dated March 12, 1987 and requested that the NRC staff reassess the violation based on the information provided in the response. The inspector reviewed the response and noted that no information was pro-vided which demonstrated that the actions taken following the testing of the "B" core spray pump on October 7, and November 25, 1987 met the requirements of Subsection IWP 3230 of the Section XI Code, Winter Ad-denda, 1980 Edition. In his response, the licensee presented his inter-pretation of a method of implementation of the code requirements; however, the licensee's position is inconsistent with the NRC staff position as established in a March 17, 1980 memorandum from IE:HQ, as described on Page 10 of Inspection Report 86-25. This. item remains open pending fur-ther review of the licensee's position to determine what additional actions may be necessary beyond those documented in FVY 87-3 .0 Observations of Physical Security  ;

Selected aspects of plant physical security were reviewed during regular and backshift hours to verify that controls were in accordance with the security plan and approved procedures. This review included the following security measures: guard staffing; vital and protected area barrier integrity; main-tenance of isolation zones; and, implementation of access controls, including authorization, badging, escorting, and searches. No inadequacies were iden-tifie I

-- - _

_ _ .

-- -

!

.

i

,

i 4.1 Security Events l

The inspector reviewed the licensee's response actions for events that occurred on March 17, 20 and 22, 1987 that involved the loss of security effectiveness. Prompt and adequate compensatory measures were taken

-

during each event and maintained for the duration of the hardware outag The events were reported to the NRC Duty officer within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per 10 j CFR 73.7 The hardware problems for the March 20th and 22nd events were

! correcte No specific cause for the March 17th failure was identified,

!- but the licensee concluded that the March 17th problem most likely oc-j curred because of the same hardware failures corrected for the March 20th i

and 22nd events. No inadequacies were identifie The inspector reviewed Security Event Repcets Nos. 87-03 dated March 23, 1987, and 87-04 and 05 dated March 25, 1987, which described the above

events. The event report accurately described the circumstances of the incident and the licensee's followup actions. Report 87-03 was submitted beyond the 5-day time period allowed by 10 CFR 73.71, based on concur-rence from the resident inspector given during a conversation with the Security Supervisor on March 20, 1987, when it appeared that the events of March 17 and March 20 were causally linked. This was subsequently determined to not be the case. No inadequacies were identifie l The inspector reviewed the licensee's plans to complete security hardware upgrades during the period of March - April, 1987, and the compensatory

measures that will be established in advance of the planned activities to assure security effectiveness is maintained. The plans and compensa-tory measures needed to complete the upgrade were described in a memor-andum to the Security Supervisor dated March 23, 1987. No inadequacies

, were identifie ,

4.2 Potential Threat Resolution The inspector received notification on March 25, 1987 from NRC Region I that, as a result of an investigation completed by federal law enforce-ment authorities, the potential security threat identified on February 3, 1987 was not considered credible (reference - Inspection Report 87-02,

'

section 4.0). This information was given to site management and security personnel on March 25, 1987. After consideration of this development,

the licensee ceased the augmented security measures established on February 3, 1987. No inadequacies were identifie i

5.0 Operational Status Reviews i

Plant tours were conducted routinely to review activities in progress and to l verify compliance with regulatory and administrative requirements. Tours of I

accessible plant areas included the control room, reactor building, turbine building, diesel generator rooms, and the protected area. Radiation controls i

'

were reviewed in areas toured to verify access control barriers, postings and radiological controls were appropriate. Plant housekeeping conditions and

J

-,-n- - , . - ,

,-,--,p g - --,,--,--y-----y,,,,-,-n-

., , .

,,----p-- - . . _ . ,-,.--..m,-n ,, - .

eg.,4c--,.-,--nag,-n,-,. , - , - - - - - - - - - - n- .w

- .A w- - - - - -

I

.

i 1 - 5

i,

)

shift staffing were reviewed. Shift logs and records were reviewed to deter-mine the status of plant conditions and the changes in operational statu Potential Reportable Occurrence reports 87-01 through 87-06 were reviewed to

, verify the events described in each were properly dispositioned by the licen-

-

see as not reportabl Actions taken by plant personnel during periods when equipment was inoperable

were reviewed to verify: technical specification limits were met; alternate

.

'

surveillance testing was completed satisfactorily; and, equipment return to service upon completion of repairs was proper. This review was completed for the following items: (1) "D" main steam line flow indication on March 17, 1987;

i "A" Main Steam Line Radiation Monitor (17-251A) failure at 4:58 p.m. on March 20, 1987. Items that received additional review are discussed further belo .1 Safety System Review

.

l The residual heat removal, core spray, residual heat removal service

water, high pressure coolant injection, service water, reactor core isolation cooling, standby liquid control and standby gas treatment systems were reviewed to verify the systems were properly aligned and fully operational in the standby mode. The review included verification that (i) accessible major flow path valves were correctly positioned, (ii) power supplies were energized, (iii) lubrication and component cooling was proper, and (iv) components were operable based on a visual inspection of equipment for leakage and general conditions. No inade-quacies were identifie .2 Feedwater Leak Detection System Status

The inspector reviewed the feedwater leakage detection system and the monthly performance summary provided by the licensee in accordance with

<

letter FVY 82-105. The licensee reported that based on the leakage monitoring data reported as of March 20, 1987, there were no deviations in excess of 0.10 from the steady state value of normalized thermocouple :

readings, and no failures in the 16 thermocouples installed on the 4 feedwater nozzles. No unacceptable conditions were identified.

>

j 5.3 Inoperable Main Steam Flow Instrument

!

!

The licensee removed differential pressure switch (dPIS) 2-119B (one of four) on the "D" main steam line at 0800 on March 17, 1987 to complete calibrations and adjustments to correct discrepancies in its readings.

, dPIS 2-1198 readings were found about 5 to 7 psi below the valves indi-j cated on the other 3 pressure switches for steam line "D". The pressure

switches are used to provide high main steam line flow isolations in the j PCIS. The operators entered the action statement for Technical Specifi-

) caticn 3.2.B and instituted a trip on the "B" PCIS train. The technical

specifications allow operation indefinitely with one inoperable channel q as long as the PCIS trip is in effect. The plant remained in the action 1 j statement until March 18, 1987 as Instrument and Control personnel in-

-- ---. -. .

- - - - - - . - -. -

._ _ _ _ __ . __ . - _ .

.

!

-

1 vestigated the switch and its sensing lines from the steamline. Actions were completed to repair the channel by installing a rebuilt switc The operators restored the PCIS to a normal standby status at 1:58 on March 18, 1987 following satisfactory completion of a functional test on the channel per OP 432 The inspector reviewed the post repair valve lineup configuration at the instrument rack to verify all main steam line channels were properly re-

stored to service, and noted that the readings on all four instrument i channels for main steam line "0" indicated about 60 psid and agreed

!

within about 5 psi. The resident inspector review of the licensee's corrective actions and maintenance activities are discussed further in

. Section 9.0 below. No inadequacies were identifie .4 Kerosene in the Connecticut River j The licensee notified the inspector at 4:30 p.m. on March 17, 1987 of

, an event that occurred offsite that was evaluated to have no impact on

! plant operations.

i j The State of Vermont notified the licensee at 3:00 p.m. that kerosene j spilled into the river upstream of the plant at 1:30 p.m. on March 17, 1 1987. About 8000 gallons of kerosene spilled into the White River in i~

the Town of Royalton, Vermont, at a location just above the confluence of the White and Connecticut Rivers. The plant is located on the Con-necticut River about 70 miles downstream of the spill site.

i The licensee evaluated the event and determined that the spill would not

,

'

have any impact on plant operations. The White and Connecticut Rivers were mostly covered with ice, and it was expected that the kerosene would l coagulate on the surface of the waters and be stopped in the ice floes.

The licensee further determined that should the kerosene ar' ve at the plant, no adverse effect on plant systems would be expected due to dilu-tion in the rivers. No further actions were planned or taken. No in-l adequacies were identified.

l

'

5.5 Failed Fuel Indications

Radiation levels increased in the offgas system on March 18, 1987. The i inspector interviewed chemistry personnel, reviewed control room strip i charts and radiation monitors, and reviewed logs and records to indepen-l dently assess the increased radiation levels and the significance of the l indications. The increased radiation levels were indicative of failed

fuel. The reactor had been operating at full power prior to the event.

!

! The daily offgas sample taken at 8:17 a.m. on March 18, 1987 showed a j reading of 1220 uCi/sec, which was up from 632 uCi/sec on March 17, 1987.

j Offgas release rates had been measured in the range of about 650 to 700

uCi/sec based on the daily samples. The results from confirmatory offgas

samples at 9:48 a.m. on March 18, 1987 verified the increases resulted l

I

_ _ - _ _ _ _ _ -

.

from activity in the offgas stream. The release rate increased to 2000 uCi/sec on March 20, 1987 and increased further to 3000 uCi/sec on April 1, 1987. The release rate was at 3360 uCi at the end of the inspection period on April 6, 1987 and appeared to have leveled off. The radiation levels at the steam jet air ejector as monitored by detectors17-150 A&B showed corresponding increases from about 40 to 70 mR/hr. The radiation levels at the inlet to the advanced offgas system guard bed increased from 1000 cpm to 10000 cp The offgas release rates remained well below the Technical Specification 3.8.K.1 limit of 0.16 Ci/sec. Reactor vessel isotopics for dose equiva-lent I-131 remain well below the Technical Specification 3.6.B.1.a limit of 1.1 uC1/gm at 1.9E-4 uCi/gm. Environmental release rates, calculated by the licensee based on the methodologies in the Offsite Dose Calcula-tion Manual, were at 0.06 mrem /yr and remained well below the Technical Specification 3.8.E.1 limits of 500 mrem /yr (whole body) and 3000 mrem /yr (skin). There were no measurable increases in the stack noble gas re-leases, based on a review of the readouts from the stack gas monitors, SG Channel I&II.

,

The increases on offgas activity occurred during routine steady state operations. The plant continued routine steady state operations after March 18, 1987, except for periods of routine surveillance testing. Ad-ditionally, plant load was reduced to 22% full power on April 4, 1987 to repair a turbine system steam leak. The increases in offgas activity did not correlate with power changes. There were large margins to core thermal limits prior to and during periods when the increases occurre The licensee's evaluation of the event was in progress at the conclusion of the inspection. Based on an evaluation of the level of activity, the isotopes present and the slope of the isotopic mixture measured from the offgas sample, the licensee's preliminary conclusion was that an equilib-

rium release mechanism existed, indicative of a pinhole type defect in one fuel rod of one fuel bundle. Possible causes for the failure include random pellet clad-interaction failure and crud induced localized corro-sion. Preliminary information from the NSSS vendor indicated that two

,

rods in fuel bundle LY 4844 (core location 29-22) came from the same rod

<

lot in use at the KKM plant in Switzerland, where indications of crud induced localized corrosion have been observed. However, fuel bundle LY 4844 is adjacent to a rodded location where the control rod went from position 14 to position 48 during the control rod pattern exchange on April 4, 1987. Measurements taken during the control rod pattern change to establish the location of the leaking bundle using the offgas flux tilt monitor did not show correlation with bundle LY 4844. The flux tilt monitor did show increases when rods on the opposite half of the core were moved. The licensee is considering plans to complete fuel sipping operations during the refueling outage to locate the defect.

The licensee's actions will be followed as part of subsequent routine inspections, along with continued reviews of plant status to confirm continued operation within license requirement ,_ .__ _

_ . _ _ _ . _ _ .

_ _ _ _ _ .

.

5.6 Turbine Control Valve Steam Leak and Power Reduction An auxiliary operator identified a minor steam leak from the above-seat drain line area of the #2 turbine control valve during routine rounds at 4:00 p.m. on April 3, 198 The reactor was operating at 100% full power at the time. Offgas release rates as measured at the steam jet air ejector remained at about 3000 uCi/sec from the failed fuel pin (s);

however, there was no detectable increase in turbine building or offsite releases as a result of the leak. Airborne radioactivity in the vicinity of the control valves remained low at about 4.9XE-10 uCi/cc. Plans were made to investigate and repair the leak during the power reduction scheduled on April 4, 1987 for routine surveillance testin ;

Following the completion of scheduled surveillance testing on April 4, 1987, reactor power was reduced to 22% full power and the turbine genera-tor was removed from the electrical grid at 5:30 p.m. A weld leak was identified on the 1 inch carbon steel drain line at the connection point to the control valve body, which was repaired by 9:45 p.m. The reactor remained at about 22% power during the repair using the turbine bypass valves for heat rejection to the condenser. The turbine was returned to the electrical grid at 10:49 p.m. on April 4, 1987 and power operation resumed under fuel preconditioning limits. The reactor was at about 85%

full power at 8:00 a.m. on April 5, 198 Oscillations in feedwater flow (1X10+6 million pounds per hour) and tur-bine pressure (30 psi) were experienced during the operation at reduced power due to control system problems. Plant control system responses returned to normal at higher powers when in three element feedwater flow control and with the turbine electrical pressure regulator in control.

-

The licensee intends to investigate suspected problems with the turbine mechanical pressure regulator and the single element feedwater flow con-

. trol mode at a later time per maintenance request (MR) 87-0659. No in-

! adequacies were identifie The inspector reviewed the status of the steam leak on April 4,1987 and the post repair conditions of the control valves following the weld re-pair. The requirements of radiation work permit 87-212 were reviewed and verified to be appropriate for the radiological hazards. The in-spector verified further that the requirements of the permit were fol-lowed during a tour of the control valve area. Control room activities and the plant status were reviewed to verify: surveillance was completed in accordance with approved procedures; plant staffing was appropriate for surveillance and plant power maneuvers; plant operators completed the load reduction in accordance with OP 0102, Power Operations (Maneu-vering at Power), Revision 13, and OP 0110, Shutdown to Low Power Standby, Revision 7; reactor coolant and offgas sampling was completed as required, and activity levels and offgas release rates were verified to remain within acceptable levels after the completion of a control pattern ad-justment; and, reporting requirements per 10 CFR 50.72(b)(2)(v) were me No inadequacies were identified.

i

-- -- , , . - .-. . - . , . , - - . - -. - -. - - - - - -

_ .. _

.

-

5.7 Plant Equipment Control The control of equipment released from service was reviewed during the inspection to verify the equipment was controlled in accordance with the ,

requirements of AP 0140, VY Local Control Switching Rule. The status of equipment associated with Switching & Tagging Orders87-230, 87-206 and 86-1454 were verified to be in accordance with the requirements of the tagging orders. The status of the equipment was determined to be J

proper for the operating condition of the facilit No inadequacies were

!

identified.

!

!

5.8 Storage Pit Wall Weepage

,

During inspection tours of the Reactor Building, the inspector noted small amounts of water weepage through the West Concrete wall of the dryer-separator storage pit. Water and sediment that had leeched out of the concrete were evident on the outside of the wall above the 318 foot elevation adjacent to the reactor building closed cooling water surge tank. The source of the water was not apparent since the storage pit has been drained since the end of the 1986 refueling outage. During a discussion with the Technical Support Superintendent on April 7, 1987, the inspector requested the licensee to review the status of the subject wall to identify the source and significance of the observed weepage, and to determine the impact (if any) on plant equipment, structures and operation.

i This item will be review further on a subsequent inspection to determine the results of the licensee's evaluation and the nature of any corrective

-

action deemed necessary, i

6.0 Organizational Changes

!

l' The licensee announced that Mr. T. Linn was selected to fill the position of

!

Security Supervisor, starting on April 20, 1987. The inspector met with the Plant Manager on March 5, 1987 to review proposed changes to the Chemistry and Health Physics Department that will result in separating the chemistry

'

and health physics functions into two groups. Under the proposal, three supervisors for the radiation protection, chemistry and operations areas would i

report to the Operations Superintendent. The plant health physicist would report to the Radiation Protection Supervisor. However, the radwaste (trans-

• portation), dosimetry and ALARA functions would be perfonned by separate groups with supervisors reporting to the Radiation Protection Supervisor.

' The emergency planning coordinator is not incorporated in the proposed organi-zation, but the position would be provided for upon completion of a decision whether to incorporate the functions in existing plant positions or a proposed new position in the Operations Support Manager staff at the corporate offic The licensee stated that he has sufficient staff to fill all positions in the

! proposed organizatio The proposed health physics organization will include

!

provisions for rotating shift coverage by individuals qualified in health i

>

,._m. y--,- -. . - ,- . . . _ . - . _ , _ . ~ . , . ..--._~,._____.-._-.m.,.,y,, . . , - , - - - . , - _ - , . , . - - - _ _ _ , , - - - . - _ . - - - . - . - . - _ . _ _ _ _-.--.c-- ---.,-- ...i

.

-

physics matters per Technical Specification 6.1.D.1, and qualified to perform limited chemistej functions to meet the " staffing for emergencies" require-ments of NUREG 073 The licensee stated that a proposed change to Technical Specification would be submitted by mid-April for NRC staff review and approval. The lic-ensee stated his intention to implement the new organization in April, 1987, pending concurrence by the NRC staff. This matter was referred to NRC Region I management and the NRR Project Manage In the existing and proposed organization, the plant health physicist is qualified to the requirements of Regulatory Guide 1.8, Revision 1 (September 1975), and would be the designated radiation protection manager. During a review of the organization with the Operations Superintendent on March 13, 1987, the licensee informed the inspector that the requirements of Specifica-tion 6.1.D.5 were inadvertently changed due to an administrative error while processing Amendment #87. The specification initially required in Amendment

1. 79 that either the plant health physicist or the radiation protection super-visor would meet the Regulatory Guide 1.8 requirements. This requirement was inadvertently changed in Amendment #87 to state that the requirement would be met only by the plant health physicist. The licensee stated he intended to maintain the flexibility provided in Amendment #79, and that thS item would be corrected in a subsequent change to the technical specifications.

'

This item is unresolved pending receipt of the proposed technical specifica-tion change from the licensee, and subsequent review and approval by the NRC staff (UNR 50-271/87-06-01).

7.0 Containment Requirements During RHR Pump Wear Ring Inspections The licensee informed the NRC orally on February 27, 1987 and by letter FVY 87-29 dated March 3, 1987 of the need to request relief from the requirements of Technical Specification 4.7. A.3 in order to perform the RHR pump wear ring inspection and replacement in a safe and efficient manner. The RHR wear ring work was scheduled to begin March 2, 1987, but was deferred pending resolution of the issue. The RHR wear ring inspection efforts and schedule has been reviewed by the NRC staff as Inspection Item 85-40-0 Technical Specification 4.7.A.3 states: " Prior to violating the integrity of a system outside the primary containment which is connected to any valve listed in Table 4.7.2.B, the isolation valves bounding the opening shall have Type C tests performed. If the opening cannot be isolated from the contain-ment by two isolation valves which meet the acceptance criteria of App'endix J (10 CFR Part 50), a blank flange shall be installed on the openin Operations staff review of the technical specification requirements during the week of February 15, 1987 determined that valves RHR 17 and ' 'R 18 in the shutdown cooling line to the RHR pumps would be used as containnent. boundary valves during the RHR work and would be subject to the requirements of the specification. Additionally, the RHR 13 valves in the torus suction lines would also be used as a containment boundary valves and are normally exempted

._- .-. . . . __

_ _ _ _ _ _ . _ - - . - - -.

. - _ . . - _ _ . ..

.

,

.

t i

from Type C leak rate testing when the RHR piping boundary forms a closed loop

!

with the containment and primary coolant system. Subsequent licensee review concluded that an acceptable Type C test could not be performed for the sub-i ject valves, but that alternative measures and controls could be established to assure containment integrity is assured in a manner comparable to Type C

!

testin The licensee's relief request and the bases for his determinations that the

alternatives to Type C testing would provide comparable assurance of accept-able containment performance during RHR pump inspections were provided in the i March 3, 1987 letter. The licensee's proposal was reviewed by the inspector, NRC technical and management staff in Region I and NRR managemen By letter i

dated March 12, 1987, the NRC Region I Regional Administrator exercised en-

!

'

forcement discretion and granted the licensee relief from requirements of Technical Specification 4.7.A.3 on a one-time basis to perform the RHR in-spections. The NRC staff safety evaluation report attached to the March 12th letter provided the basis for the action, that was subject to the conditions

!

described below which were in addition to those stipulated in the licensee's March 3rd submittal. A copy of the proposed conditions, described below, were j provided to the licensee staff.

During discussions between the resident inspector and the plant staff during the period of March 3-9, 1987, it was determined that the licensee shall:

j (1) Administratively control the V10-13A, B, C and D and the V10-15A, B, C and D valves as follows: locked closed with chain and lock, along with opening

,

their MOV supply breaker, and subsequently test that the proper supply breakers are open by attempting to open each valve using the CRP 9-3 control switch, one at a time. (2) Provide an on-line local pressure indicator in addition to the PS-118 pressure switch, both of which shall be surveilled once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to provide early indication of any leakage through the X12 penetratio (3) Establish a hold point in the pump inspection procedure that will verify, after pump drain down and prior to RHR pump disassembly, that there is no measurable leakage from penetrations X12, X224A, and X224B, i.e., no steady stream. (4) Maintain surveillance over the open boundary using operator tours once every four hours. If measurable leakage from the containment is detected, i.e., a steady stream, the shift supervisor shall order the pump reassembled expeditiously. Should a need arise to implement this action, the RHR pump j reassembly shall be performed on a round-the-clock basis, with the intent to

! restore the pump to an operable status with a new impeller. (5) Develop and j implement a contingency plan (using the pump reassembly procedure) that will provide for rapid closure of the RHR piping boundary (e.g., in about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />)

in response to a decision by the operations shift supervisor that such action is necessary, if leakage conditions do not permit reassembly using the new impelle Based on the Staff's evaluation and the licensee's detailed plans and controls j for repairing the pumps, the NRC staff concluded that the intent of Technical

,

Specification 4.7.A.3 would be met during the performance of the activity, i The inspector had no further comments on this area at the present time. The

! start of RHR pump inspections was subsequently rescheduled for the week of i

!

.- - ,_ , _ _ _ . _ _ _ _ _ ._ - _ _ _ ___._ _ __ _ _ . - _ _ _ ,

-

_m. . . _ _ _ _ _ . _ - _ _ _ _ , _ _ _ _ _ _

-- . __ ________ ___

.

April 20, 1987 due to the need to upgrade procedures to address the above mentioned conditions, and due to the availability of the primary contractor and his workers. The RHR repair effort and licensee conformance with the established procedural controls and the conditions stipulated in the NRC's March 12, 1987 letter, will be reviewed further on subsequent routine inspec-tions. No inadequacies were identifie .0 Proposed Spent Fuel Pool Changes The inspector began a review during this inspection period of Engineering Design Change Request (EDCR)87-405, along with its references and enclosures, that was prepared by the licensee for proposed modifications to the spent fuel pool at the facilit The licensee concluded in the design change package dated March 27, 1987, that the facility could be changed in the manner de-scribed without creating an unreviewed safety question as defined by 10 CFR 50.5 The EDCR contains a safety evaluation (Section 15.0) that provides the licen-see's basis to install a single Nuclear Energy Services (NES) high density storage rack with an 18X20 array in the pool. The installation of an addi-tional 360 storage cells will increase the pool capacity to 2050 spent fuel bundles, which is more capacity than the current license limit of 2000 set by Amendment #37. However, the technical specification limit is based on stored spent fuel bundles and not storage capacity. While the new storage rack is not needed to accommodate the 136 fuel bundles scheduled to be dis-

! charged from the reactor during the refueling outage starting in August, 1987, it is assumed that the new rack would be used for a future core off-load if that were needed. The only preparatory work required to install the rack will be to relocate spent fuel already in the pool to empty two Programmed & Remote, Inc. (PAR) type racks that are presently in the spent fuel pool. PAR racks

1. 15 and #18 will then be relocated within the pool to make room for the NES rack. The work is scheduled to be done starting the week of May 25, 198 This matter was referred to NRC Region I to request assistance in completing i

a technical evaluation of the proposed change to the facility, and to deter-mine specifically whether (1) the change is of a type that can be processed under 10 CFR 50.59; and, (2) the change creates an unreviewed safety question as defined in 10 CFR 50.59(a)(2). This item is unresolved peiding completion of the NRC staff review of the licensee's proposed actions (UNR 50-271/87-06-02).

9.0 Maintenance Activities Maintenance activity was reviewed to determine the scope and nature of work l done on safety related equipment and equipment referenced in the technical 1 specifications. The review confirmed: the repair of safety related equipment '

received priority attention; technical specification limiting conditions for operation were met while components were out of service; performance of al-ternate safety related systems was not impaired; and, the activity was com- j pleted in accordance with established procedures and plant equipment control j l

. _ , . . _ . _ _ . . , _ _ _ __ - _ _ _ _ _ _ . . . _ . . _ . . - _ _ _ _ _ - _ _ _ . _ _ , . ,

. - . _ _ . - - . _ _ _ --

i *

I i

!

i

,

' Maintenance activity associated with the following was reviewed to verify (where applicable) procedure compliance and equipment return to service, in-cluding operability testin MR 87-0398, "A" Main Steam Line Monitor Inoperable on 3/20/87

--

MR 87-0047, dPIS 2-1198 Inoperable on 3/17/87 No inadequacies were identified, except as noted below.

,

9.1 Main Steam Line Flow Instruments, MR 87-0047 Differential pressure switch dPIS2-119b was removed from service on March j 17, 1987 to investigate a non-linearity at the mid point of the instru-

ments range while on-line. The instrument read about 5 psi below the j

i other three instruments on the same steam line. The instrument had tested satisfactorily during previous functional and calibration check In addition to further functional and calibration testing on March 17, 1987, the licensee flushed and inspected the instrument sensing line snubbers, performed a leak check and measured the static zero point on the instrument at 1000 psig, in an attempt to identify the cause for the i

'

non-linearity. The licensee concluded that the differential pressure unit on the transmitter was faulty and that it would respond in a non-j linear manner only under full system operating pressure of 1000 psig.

.

Actions were taken to replace the faulty unit on March 17, 1987 with a i new unit from stores. However, the new unit was removed from the system

!

after it was found to be defective when it was installed and functionally 4 tested. A third differential pressure switch was assembled by I&C per-j sonnel using spare units, by combining a differential pressure unit from

>

one instrument with a movement assembly from another unit. The rebuilt

j switch was installed, tested satisfactorily and returned to servic The inspector verified that all four flow. instruments were returned to i service and that the instrument readings agreed within 5 psi upon com-j pletion of the maintenance activity. No inadequacies were identifie .0 0QG Followup of Audit Findings The YNSD Director of Quality Assurance informed the NRC inspector on April

! 3, 1987 regarding a quality assurance (QA) inspector formerly assigned to the j VY site with the Operational Quality Group (0QG), who left the employ of YNSD on March 27, 1987 purportedly due to a personal problem with another indivi-

'

i

'

dual at the sit The licensee stated that the QA inspector stated his in-tentions to contact the NRC to express concerns with how the onsite 0QG per-

forms its functions, j

' The licensee provided information relating to known concerns expressed by the QA inspector, which were raised as issues during the Summer of 1986. The licensee stated that actions were taken in an attempt to address the concerns.

,

'

The licensee stated that his investigation of the individual's concerns found i

the following: (1) There is disagreement that persists between the QA inspec-i i

-. - - - - - - - - _ . - _ - _ - ._- - -.. .- -_. -_- ,-_.,-._ -_.-_-- - -_ -_., -

i

.

,

I tor and QA management on how to complete programmatic closeout reviews of QA

audit findings. The 0QG relies to a great extent on " paper" reviews instead of inplant inspections with indepth followup. The licensee feels the present

'

system is satisfactory since subsequent QA audits reinspect the areas where deficiencies were found to assure corrective actions were' successful. The QA inspector identified no specific examples wherein " paper" review closeout of open items resulted in an unsatisfactory followup of audit findings. There are no known problems that would affect the material plant or cause a safety hazard. (2) There was a personality conflict with another individual. The previous onsite 0QG supervision did not correct the problem. The licensee stated that he believes the current 0QG organization would address any sub-sequent concerns in this area. (3) Personnel performance problems were iden-

tified that require licensee management attention for correction. The nature of the problems were categorized by the NRC inspector as a " licensee internal matter".

The NRC inspector requested that the licensee have the QA inspector contact the Region I Office or the VY Resident Inspector Office should he contact the licensee again so that his concerns can be further reviewe This item is unresolved pending further NRC review of the 0QG functions to assure Quality Assurance program requirements are being met, and to assure the QA inspector's concerns are fully resolved (UNR 50-271/87-06-03).

i'

11.0 Attentiveness to Licensed Duties l The inspector received a press release from NRC Region I dated March 31, 1987 regarding recent inspection findings at the Peach Bottom facility concerning licensed operator inattentiveness and the subsequent NRC action to shutdown

.

the plan The press release was provided to the licensee's staff (including ( the President, the Plant Manager, the Operations Supervisor and control room personnel) for information and review. Additionally, the inspector discussed i

'

the inspection findings at Peach Bottom and its significance to the NRC staff with control room personnel on the 4:00 p.m. to midnight shift on April 1,

198 l

The inspector noted that no instances of inattentiveness to licensed duties have been observed during routine and backshift inspection at Vermont Yankee, based on inspector observations of control room activities during " unannounced" entrances into the room, and an operator awareness of plant status indicators as displayed on control room recorders and indicatons. Future NRC actions relative to operator attentiveness will be followed by the inspector during subsequent routine inspection No inadequacies were identified.

12.0 Surveillance Activities l The surveillance test results listed below were reviewed to verify that test-

ing was performed in accordance with approved procedures, test results demon-strated compliance with technical specification and administrative require-ments, and that deficiencies (if any) were corrected in accordance with es-tablished administrative requirements.

l

_ - -. ._. ., _ _ ._ . _ ,.. _ _ ._ . . - _ _ _ _ _ _ _ __ _ _ . . . _ . , _ - - _ , - _ _ _ . . _ _ _ _ . _ _ _ _

.

+ OP 4511, Main Steam Line Monitor Setpoint Check, reviewed 3/13/87

+ OP 4323, Main Steam Line High Flow Functional Test, reviewed 3/17/87 No inadequacies were identified, except as noted belo The items discussed below warranted inspector followu .1 Instrument Setpoint Changes During a review of alarm and trip setpoints for the main steam line radiation monitors, the inspector noted that the surveillance results recorded on VYOPF 4511.17 for testing completed on February 17, 1987 indicated that the alarm setpoint for the radiation monitor on the "C" main steam line was out of specification. This area was previously re-viewed as Inspectior Item 87-04-01, as noted in Section 3.0 above. Based on the February 17tt test results, a setpoint change request (SPCR 87-02)

for the "C" monitor was initiated per AP 0022 to change the limits from 198 mR/hr to 180 mR/h However, as of March 13, 1987, no action had been taken to complete the setpoint change reques This item was discussed with Instrument and Control (I&C) and Chemistry &

Health Physics (C&HP) personnel. The I&C Department was responsible for implementing the request even though the setpoint change was initiated by C&HP. The inspector determined that SPCR 87-02 was not implemented since the channel was subsequently found to be within specification on March 1, 1987, and completion of the request would have placed the moni-tor out of calibration. However, this action and its basis was not tracked in the only status log maintained for setpoint change request Some confusion existed for licensee personnel on March 13, 1987 regarding the status of the request because of the inadequacies in the status lo The licensee stated that a recent QC inspection had identified the same discrepancy with the log, along with other problems with the controls as established in AP 0022. The licensee provided P i nspection Report I&C 87-07 dated February 20, 1987 for inspector - The inspector noted that the QC report identified several defic.s .es in AP 0022 that required correction to improve programmatic control of setpoint change The inspector noted further that none of the deficiencies involved dis-crepancies with the material plant. This item will be reviewed further on a subsequent inspection pending completion of the licensee's review and dispositioning of the discrepancies identified in QCIR I&C 87-07.

i

'

The inspector noted that the alarm setpoint for the monitors were proper for the full power background conditions on March 13, 1987. The inspec-tor expressed his concerns regarding the length of time from February

,~ 17th until March 1 when the alarm setpoint could have been outside the range required by administrative procedures and the technical specifica-tions. The licensee stated that problems have been experienced in main-

,

'

taining the proper alarm setpoints for the channels due to: drift on the instrument channel drawers; variations in full power background readings; the need to maintain relatively tight setpoint tolerances on an instru-ment channel with a design accuracy specification of +/- 25% for the

_ __ _ . _ __ _ _ _ . _ _ . _ _ _

.

.

+

'

.

operating range of interest (reference GEK 32426); and, a design that uses a two pen recorder that can only select 2 of 4 channels for the display and alarm functions at any one time. Subsequent to the inspec-tor's review, the licensee reviewed the performance of the 4 main steam line channels, and implemented setpoint change requests 87-03 on March 13, 1987 to raise the B/D high alarm setpoint, and 87-04 and 87-05 on March 17, 1987 to readjust the high and high-high setpoints on all four monitors, due to changes in background readings. The licensee initiated additional reviews to determine whether improvements in calibration practices could be made that would either provide more reliable alarm setpoints or otherwise address setpoint " drift" problems. This item will

be reviewed further on a subsequent inspectio Based on the above, the inspector had no further questions at the present regarding the operability of the channels and the adequacies of the ex-isting surveillance and testing procedures to detect and correct out of specification conditions. This item is unresolved pending completion j of licensee actions as noted above and subsequent review by the NRC (UNR

! 50-271/87-06-04).

I

! 13.0 Shipment of Irradiated Materials From the Spent Fuel Pool i

The inspector reviewed activities during the inspection period to remove ir-

! radiated components from the spent fuel pool and ship them to Hanford, Wash-

, ington for burial. The radioactive waste was consigned to U.S. Ecology's Hanford Reservation and was a controlled route shipment designated as VY Shipment No. 87-9. Notification of the shipment dates and the shipping routes d

was provided to the appropriate states. Advanced notification was made to

NRC Region I by letter VYV 87-96 dated March 31, 1987.

a lhe irradiated components consisted of used reactor dry tubes, local power

"ange monitors, and source range monitors (fission chambers), packaged in

three liners. The reactor waste contained a total of 0.171 grams of special nuclear material, and a total of 11,596 curies of activity, attributable aostly to the isotopes of Co-60, Fe-55, Ni-63 and Mn-54 present in the waste

'

,

t.s activation and corrosion products. The liners were shipped in a Trans-nuclear TN-8L spent fuel shipping cask marked USA /9015/B()F, which was de-j scribed in a NRC Certificate of Compliance dated May 12, 1986. The highest j dose rates measured on the liners was 24,000 R/hr, and the highest contact ;

! dose rate on the loaded shipping cask was 120 mrem /hr. The highest dose rate on contact with the closed, exclusive use vehicle was 10 mrem /hr.

i The inspector reviewed licensee procedures, administrative controls and the i

activities in progress on March 23-27, 1987 to load the TN-8L cask. Cask handling operations in and around the spent fuel pool were also witnessed by l the inspector. The inspection included verification and/or consideration of:

i NRC approval of the cask lifting device; certification of lifting device load l testing; requirements and controls established by procedure OP 2208.3, Revi-I sion 0 with DI 87-06 dated March 23, 1987, including completion of cask pre-i service tests and surveillance; controls to mechanically and administratively i,

i i

_

'

c

-

restrict cask movement around and over the spent fuel pool; completion of prerequisites per OP 2208.3; maintenance crew training and pre-job briefings; inclusion of QC hold points and QA surveillance by the onsite operational quality group; compliance with technical specification 3/4.12.G requirements; and, health physics controls for work on the 345 foot elevation of the Reactor Building and around the spent fuel poo The inspector reviewed licensee plans and preparations to ship the waste from the site on March 31, 1987. The cargo vehicle was labeled Radioactive - Yel-low III with a Transport Index of 8.0. The inspection included verification and/or consideration of: liner and cask radiation surveys to meet the re-

quirements of Certificate of Compliance 9015 dated May 12, 1986; package and

vehicle dose rate conformance with the limits of 49 CFR 173.441; waste clas-

,

i sification per the requirements of 10 CFR Part 61; package decay heat loads for conformance with the requirements of the Safety Analysis Report dated

Decemoer 27, 1985, the Certificate of Compliance dated May 12, 1986, and 49 i CFR 173.442; and, determination of activity limits for nuclides per the re-

] quirements of 49 CFR 173.43 The waste was shipped from the site on March 31, 1987 and arrived at the burial site on April 5, 1987. No inadequacies were identifie .0 Management Meetings

Preliminary inspection findings were discussed with licensee management peri-

'

odically during the inspection. A summary of findings for the report period was also discussed at the conclusion of the inspection and prior to report issuanc :

$

i

,

I l

l

- .. _ .. . . _ _ _ . _ . - _ . _ . . _ _ - - . . - - - -. __..- _

_ _ _ - _ - - _ _ -.