ML20236L538

Notice of Violation from Insp on 980302-0409.Violation noted:post-mod Test Procedure Did Not Incorporate Requirements Contained in Design Documents

ML20236L538
Person / Time
Site:	Millstone
Issue date:	06/11/1998
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20236L533	List: IR 05000336/1998202 ML20236L532 ML20236L538
References
50-336-98-202, EA-98-271, NUDOCS 9807130063
Download: ML20236L538 (5)

Text

-_ _ - _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ ____ _ -

NOTICE OF VIOLATION Northeast Nuclear Energy Company Docket No. 50-336 Millstone Nuclear Power Station License No. DPR-65 Unit 2 EA 98-271 During an NRC inspection conducted on March 2 through April 9,1998, violations of NRC requirements were identified. In accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions," NUREG-1600, the violations are listed below:

A. 10 CFR Part 50, Appendix B, Criterion XI, " Test Control," requires that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate j the requirements and acceptance limits contained in applicable design documents. i Contrary to the above:

1. A post-modification test procedure did not incorporate the requirements contained in design documents. Test procedure SP 21206A, " Instrument Air Accumulator Check Valve Test,"

Rev. 3, dated November 5,1997, tested the backup air accumulators for valves 2-RB-13.1A/B l

but did not incorporate the requirements and acceptance limits contained in calculation 97-ENG-01823-M2, " Verification of Accumulator Size for Valves 2-RB-13.1A and 13.18," Rev. O, dated August 13,1997, and modification PDCR 2-064-95, " Air Accumulator for 2-RB-13.1 A & B," Rev.

1, dated August 20,1997. The test did not verify that the air accumulated leak-tightness was sufficient to hold the valve in the closed position for 90 minutes with a starting pressure of 90 pounds per square inch gauge (psig) and a final pressure of 60 psig.

2. _ All testing required to demonstrate that the reactor building closed cooling water I system (RBCCW) will perform satisfactorily in service was not identified. A periodic test to verify j adequate flow to each component serviced by RBCCW had not been established. Procedure SPROC 97-2-19, "RBCCW Building Closed Cooling Water System Flow Balance," Rev. 2, dated March 2,1998, only required a one-time flow balance be performed.

3. A written test procedure did not incorporate acceptance limits. Preventive maintenance l

procedure MF 2701J-96, " Service Water Cooled Heat Exchangers Subject to GL 89-13,"

Rev. 3, dated April 21,1997, provided instructions for periodic maintenance and inspection of i the service water cooled heat exchangers. The procedure did not provide acceptance limits for the as-found cleanliness of the RBCCW heat exchangers.

This is a Severity Level IV violation (Supplement 1).

B. 10 CFR 50, Appendix B, Criterion XVI, " Corrective Action" requ' ires that conditions t

adverse to quality be promptly identified and corrected.

Procedure RP-4, " Corrective Action Program," Rev. 6, Change 1, dated April 1,1998, requires that conditions adverse to quality be identified and promptly corrected.

9907130063 980611 PDR ADOCK 05000336 G PDR I.

O Contrary to the above:

1. A condition adverse to quality had not been adequately corrected. Condition report M2-97-0489, *RBCCW System Design Pressure Can Be Exceeded at Low Flows," dated March 27,1997, stated that RBCCW system pressure could exceed design pressure during pump swapping. - Operating Procedure OP 2330A, "RBCCW System," Rev.19, dated March 9,1998, was changed to alleviate the pressure spiking. The procedure was not tested

for effectiveness at low flows and was inconsistent with Final Safety Analysis Report (FSAR)

Section 9.4.4.2 requirements and with surveillance test requirements.

2. A condition adverse to quality had not been identified nor corrected. Technical Specification (TS) 3.8.1.1 requires two sources of offsite power be supplied to the switchyard, whereas 10 CFR Part 50, " General Design Criterion 17.," Appendix A, requires that two sources be supplied to the safety buses. LER 95-035, dated October 5,1995, reported that the licensee procedures had not required them to enter a TS limiting condition for operation with less than two power paths from the switchyard to the onsite safety busses. The licensee procedures were changed at that time but the need for a TS change was not identified nor corrected.

This is a Severity Level IV violation (Supplement 1).

C. 10 CFR Part 50, Appendix B, Criterion Vlli, " identification and Control of Materials, Parts, and Components," requires that measures be established for the identification and control of materials, parts, and components, and that the identification be maintained.

Contrary to the above, the identification and control of a valve was not maintained.: In several databases, letdown heat exchanger RBCCW outlet temperature control valve, 2-RB-402, had two' different identification numbers. The valve was identified as 2-RB-402 and 2-CH-223.

Valve 2-RB-402 was identified as safety-related, whereas 2-CH-223 was not.

This is a Severity Level IV violation (Supplement I).

D.. 10 CFR Part 50, Appendix B, Criterion V, " Instructions, Procedures, and Drawings,"

states in part that activities affecting quality be prescribed by documented instructions and be accomplished in accordance with those instructions.

Contrary to the above:

1. An instance was noted where the two RBCCW pump trains (Facilities 1 and 2) did not meet the electrical separation criteria specified in FSAR Section 8.7 and the licensee specification SP-M2-EE-0016, " Electrical Separation Specification- Millstone Unit 2," Rev.1, dated September 9,1997. The separation criteria required 18 inches of free air space horizontally between redundant cable trays. For cables Z12AA20, Z2LAA20, Z24LA60, and Z16HT35 on Standards drawing 25203-34031, Rev. 7, there was approximately 9 inches of free air space horizontally.

I e

2. Two pressure indicators, PI-6324 and PI-6325, and their respective tubing, were incorrectly classified as nonsafety-related and nonseismic. FSAR section 5.2.8.2.1 and specification SP-ME-668, Rev. 4, dated May 23,1997, required these instruments and tubing to be seismic and safety-related.

l

3. An example of originally-installed small bore piping which was not installed in accordance with the applicable piping isometric drawing was identified. A 3/4-inch by a %-inch reducing elbow was installed instead of a 3/4-inch elbow and a 3/4-inch by a %-inch reducer as specified by DCN No. DM2-00-0640-97, "HPSI 41 A Seal / Bearing Coolers C . ' ling Water Supply and Retum Pipe Replacement," dated August 12,1997. Additionally, a pipin; support was installed differently than required. The support, Hanger No. 6, was shown on l drawing 25203-22200 SH.491315E, Rev. 00, dated March 3,1982. The drawing showed a frame type restraint but the installed restraint was a cantilever arrangement. Additionally, the drawing showed a 4-inch and a 2-inch angle members, but the installation was made of a 6-inch square tube and a 6-inch wide flange members. Likewise, anchor base plate locations, sizes, and concrete fastener sizes were different than shown on the drawing.

4. There were two examples of recent modification work where piping supports were not installed in accordance with their drawings. The supports,25203-22200-611087 and 25203-22200-611100, were shown on DCN Number DM2-00-0919-97 *HPSI Pump P-41A Seal / Bearing Cooling Water Supply and Return Piping Supports," Rev. O, dated October 10,1998. The actualinstallations differed from the drawing in that the assemblies were rotated 90 degrees from that shown.

5. An FSAR change request had not been initiated as required by Procedure RAC 03,

" Changes and Revisions to Final Safety Analysis Reports," Rev.0, dated December 18,1997.

The flow for the shutdown coolers described in FSAR Table 9.3-1 were not revised to reflect the reduction of the design basis minimum design flow from 4820 gallons per minute (gpm)

(2.41 x 105 pounds per hour) to 3500 gpm (approximately 1.75 x 105 pounds per hour). A request to change FSAR Table 9.3-1 had not been made.

This is a Severity Level IV violation (Supplement 1).

E. 10 CFR Part 50, Appendix B, Criterion lil, " Design Control," requires in part that measures be established to assure that the design basis is correctly translated into specifications, drawings, procedures, and instructions.

Contrary to the above, for RBCCW radiation monitor, RM 6038, the design basis alarm setpoint calculation assumptions for flow were not translated into operating procedures. OPS Form 2669A-2, " Unit 2 Aux Building Rounds," Rev. 25, did not specify the flows assumed in the setpoint calculation. Similarly, the " Millstone Two Radiation Monitor Manual," dated Jurse 27,1997, provided an inadequate setpoint calculation bases for the RM-6038 alarm setpoint in that it did not assume flow dilution. Additionally, Operations Form 2654K, "Rediation Monitor Setpoint Verification," Rev. 3, did not provide guidance for dealing with postulated conditions of high background radiation levels which could result in a nonconservative setpoint.

This is a Severity Level IV violation (Supplement 1).

F. 10 CFR Part 50, Appendix B, Criterion V, " Instructions, Procedures, And Drawings,"

requires that activities affecting quality be prescribed by procedures of a type appropriate to the circumstances.

Contrary to the above, the annunciator response procedures for the RBCCW system contained

. numerous inconsistencies, had widely differing levels of detail, and exhibited poor integration with existing operating and abnormal operating procedure instruction (AOP) 2564, " Loss Of RBCCW," Rev. 3. The problems identified are detailed in the inspection report.

l This is a Severity Level IV violation (Supplement 1).

G. 10 CFR Part 50, Appendix B, Criterion VI, " Document Control," requires that measures be established to control issuance of documents, such as instructions, procedures and drawings, including changes thereto, which prescribe activities affecting quality. These measures shall assure that documents, including changes, are distributed to and used at the prescribed location.

~ Contrary to the above, the full-size Unit 2 piping and instrumentation diagrams maintained in the Controlled Document Library on the third floor of building 475 had not been updated since May 1997.

This is a Severity Level IV violation (Supplement 1).

H. 10 CFR 50.59, " Changes, Tests, and Experiments," permits changes to be made in the facility as described in the FSAR, and requires that records of the changes be maintained, and that the records include a written safety evaluation which provides the bases for the -

determination that the change did not involve an unreviewed safety question. A proposed change involves an unreviewed safety question if the probability of occurrence or the consequences of a malfunction of equipment evaluated in the safety analysis report may be increased.

NRC " Safety Evaluation of the Millstone Point Nuclear Power Station," dated May 10,1974, Section 7.10, approved 12 inches as the electrical cable separation criteria in panels.

FSAR Section 8.7.3.1 described the same separation criteria.

Contrary to the above, a safety evaluation erroneously concluded that a reduction in the plant-wide electrical separation criteria was not an unreviewed safety question. Design Change Record (DCR) M2-96-068, " Electrical Separation Specification-Millstone Unit 2," Rev 0, dated

. September 8,1997, revised SP-M2-EE-0016, Rev. O, dated September 9,1996. The change reduced the electrical cable separation criteria from 12 to 6 inches in cabinets. The DCR included a Safety Evaluation No. S2-EV-97-0018, Rev.1, dated September 8,1997, which concluded that there was not an unreviewed safety question. The reduction in separation could increase the probability of a previously evaluated malfunction of equipment.

This is a Severity Level IV violation (Supplement 1).

O Pursuant to the provisions of 10 CFR 2.201, Nor+heast Nuclear Energy Company is hereby required to submit a written statement or explanation within 30 days of receipt of the letter transmitting this Notice of Violation to the U.S. Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington, D.C. 20555, with a copy to the Director, Special Projects Office, Office of Nuclear Reactor Regulation, and a copy to the NRC Resident inspector at the Millstone Nuclear Power Station, Unit 2. This reply should be clearly marked as a " Reply to a Notice of Violation," and should include for each violation (1) the reason for the violation, or, if i contested, the basis for disputing the violation, (2) the corrective steps that have been taken j and the results achieved, (3) the corrective steps that will be taken to avoid further violations,-

and (4) the date when full compliance will be achieved. Your response may reference or include previous docketed correspondence, if the correspondence adequately addresses the required response, if an adequate reply is not received within the required time specified in this Notice, an order or a Demand for information may be issued as to why the license should not be modified, suspended, or revoked, or why such other action as may be proper should not be taken. Where good cause is shown, consideration will be given to extending the response time.

Dated at Rockville, Maryland this 11th day of June,1998 l

l 1

___ ___ . _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ - _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ -