ML20216G092

Notice of Violation from Insp on 971027-980128.Violation Noted:Licensee Failed to Perform & Include in Written Records,Safety Evaluations for Approx 250 Changes to FSAR Made Between mid-1996 & mid-1997

ML20216G092
Person / Time
Site:	Millstone
Issue date:	04/01/1998
From:	NRC
To:
Shared Package
ML20216G083	List: IR 05000423/1997209 ML20216G078 ML20216G092
References
50-423-97-209, NUDOCS 9804170341
Download: ML20216G092 (4)

Text

.

\

i NOTICE OF VIOLATION l i

Northeast Nuclear Energy Company Docket No. 50 423 Millstone Nuclear Power Station License No. NPF-49 Unit 3 During an NRC inspection conducted from October 27,1997 through January 28,1998, violations of NRC requirements were identified. In accordanca with NUREG-1600, " General Statement of Policy and Procedure for NRC Enforcement Actions," the violations are listed below:

A. Technical Specification 6.8.1.s requires, in part, that written procedures be established -

and maintained in accordance with the applicable portions of Regulatory Guide 1.33, " Quality Assurance Program Requirements," Appendix A, Revision 2, February 1978. ' The applicable portions of Regulatory Guide 1.33, Appendix A, Revision 2, include procedures for equipment surveillance and annunciator response.

Contrary to the above, two examples were identified where the requirements of TS 6.8.1a were not met:

(1) Surveillance Procedure (SP) 344SE12, " Protection Set 1 RCS Narrow Range RTD Time Response (Red)," Revision 0, Change 2; and the corresponding procedures for the other reactor protection sets, SPs 3443E73, E30, E40, E21. E31, E22, E32, and E42, failed to include a determination of data acceptabi5ty prior to its use.

(2) Surveillance Procedure (SP) 31024, " Calculation of Reactor Trip and ESF Response Times," Revision 2, did not increase the results by 10 percent to account for the tolerance of the analysis method as suggested by the vendor manual. This resulted in a failure to include margins to account for the indirect measurement of RTD response time.

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

V B. 10 CFR 50.59(b)(1) states, in part, that the licensee shall maintain records of changes to the facility. These records must include a written safety evaluation which provides the bases for the determination that the change, test, or experiment does not involve an unreviewed safety question.

Contrary to the above, the licensee failed to perform and include in their wntten records, safety evaluations for approximately 250 changes to the FSAR made between mid-1996 and mid-1997. The licensee determined that these safety evaluations were not included in the scope of the corrective actions that resulted from their identification, in June 1997, of an inappropriate threshold for the preparation of safety evaluations.

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

C. 10 CFR 50.71 (e) requires, in part, that the licensee update the FSAR to assure that the FSAR contains the latest information developed at an interval not to exceed 24 months.

9804170341 980401 I PDR ADOCK 05000423 L G PDR  !

i '

Contrary to the above, the following examples were identified where the information contained in the FSAR had not been updated within the past 24 months:

(1) FSAR Section 6.2.2 describes the RSS pump seal as being continuously protected by a  ;

sealinjection system, thus the RSS pump seals are not expected to see the effects of particles in the pumpage. Actually, the design does not allow for long-term seal injection and the pump  ;

can operated with particles in the pumpage without significant degradation cf the seals. l (2) FSAR Sections 15.6.3.1,15.6.3.2, and 15.6.3.3, described reducing the reactor coolant system pressure to the faulted steam generator pressure in order to stop the flow through the broken tube. These FSAR sections indicate that the cooldown is achieved by the release of

)

steam from the nonfaulted steam generators' safety / relief valves which was not consistent with '

Westinghouse Ana'jses WCAP-11002, " Evaluation of Steam Generator Overfill Due to a Steam Generator Tube Rupture Accident," dated February 1986, and WCAP-10698-P-A, " Evaluation of Offsite Radiation Doses for a Steam Generator Tube Rupture Accident," dated March 1986.  !

These analyses assumed that operators would cool down the reactor coolant system by controlling the release of steam from the nonfaulted steam generators through the steam generator dump valves. l (3) The results of Calculation 88-019-96RA, "EAB [ Exclusion Area Boundary) and LPZ [ Low  !

Population Zone) Doses From a Unit 3 LOCA [ Loss-of-Coolant Accident)," Revision 2, dated November 2,1993, were not corslatent with FSAR Table 15.0-8, " Potential Offsite Doses Due to Accidents."

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

D. Technical Specification 3.1.2.1 requires that a boron injection flow path be OPERABLE and capable of being powered from an OPERABLE emergency power source when the plant is in a shut:fown condition.

Contrary to the above, from October 11 to November 15,1997, Unit 3 was in a volume control tank outage and the "A" charging pump, with its associated emergency diesel generator (EDG),  !

was required for the reactivity flow path. A failure of the "A" EDG building ventilation air intake damper on October 11,1997, rendered the "A" EDG inoperable. For 36 days, an emergency power source was not available to support the boron injection flow path.

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

' E. 10 CFR Part 50, " Instructions, Procedures, and Drawings," Appendix B, Criterion V, requires, that activities affecting quality be prescribed by and accomplished in accordance with documented instructions, procedures, or drawings of a type appropriate to the circumstances.

Contrary to the above, bypassfjumper 3-97-060, which was used to install temporary space heaters in the train "A" EDG building, was installed contrary to the instructions contained within the temporary modification. The temporary heaters were required to be installed more than 6 ,

inches above the building floor to reduce the risk of fire in the unlikely event of an oil or fuel spill. During a walkdown of the building, the NRC found they were installed approximately 4 inches above the building floor.

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

t F.

The licensee's Design Control Manual (DCM) Chapter 5, " Calculations," required that calculations (1) document the applicable references and design inputs, (2) clearly state the assumptions and their bases, (3) describe the method of calculation, (4) be design verified, and (5) be numbered and logged.  !

Contrary to the above, bypass / jumper 3-96-111, which was implemented to block open auxilia feedwater cross-connect valves was supported by an uncontrolled calculation that did not contain the information required by the DCM.

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

G. 10 CFR Part 50, Appendix B, Criterion ill, requires, in part, that design control measures be provided for verifying or checking the adequacy of design.

Contrary to the above, the appropriate pressure limits were not translated into a temporary modification to the facility in Bypass / Jumper 3-96-070, which designed a temporary cover plate i on the train 'C' main steam isolation valve.

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

Pursuant to the provisions of 10 CFR 2.201, Northeast 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 Of5ce, Office of Nuclear Reactor Regulation, and a copy to the NRC Resident inspector at the

• Millstone Nuclear Power Station, Unit 3. 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 contested, the basis for disputing the violation, (2) the corrective steps that have been taken 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 i

include previous docketed correspondence, if the correspondence adequately addresses the required response. If an adequate reply is not received within the required time specifed in this Notice of Violation, 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.

If you contest this enforcement action, you should also provide a copy of your response to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555.

Because your response will be placed in the NRC Public Document Room (PDR), to the extent possible it should not include any personal privacy, proprietary, or safeguards information so that it can be placed in the PDR without redaction. If personal privacy or proprietary information .

is necessary to provide an acceptable response, then please provide a bracketed copy of your i response that identifies the information that should be protected and a redacted copy of your  !

response that deletes such information. If you request withholding of such material, you mual specifically identify the portions of your response that you seek to have withheld and provide in detail the bases for your claim of withholding (e.g., explain why the disclosure of information will l create an unwarranted invasion of personal privacy or provide the information required by

s ,

10 CFR 2.790(b) to support a request for withholding confidential commercial or financial information). If refeguards information is necessary to provide an acceptable response, please provide the level of protection described in 10 CFR 73.21.

Dated at Rockville, Maryland this 1" day of April,1998 1

i l l l

l i l

V.

.