ML17262A795
| ML17262A795 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 03/26/1992 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17262A794 | List: |
| References | |
| 50-244-91-201, NUDOCS 9204030022 | |
| Download: ML17262A795 (8) | |
Text
ENCLOSURE 1 NOTICE OF VIOLATION Rochester Gas and Electric Corporation R. E. Ginna Nuclear Power Plant Docket No. 50-244 License No. DPR-18 During the U.S. Nuclear Regulatory Commission's (NRC's) Service Water System Operational Performance Inspection conducted between December 2-20, 1991, three violations were identified. In accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2, Appendix C, (1991), the violations are listed below:
A.
10 CFR 50, Appendix B, Criterion III, "Design Control," requires, in part, that design interface controls be established and that design control measures be provided for verifying or checking the adequacy of design.
Ginna Quality Assurance Manu@, Section 3, "Configuration Control," Rev. 13, dated November 1, 1986, states in Section 3.4.3, "Design Verification," "The design verification shall assure that the design outputs (i.e., drawings, analysis and specifications) including design outputs from equipment suppliers, meet the design input requirements and are consistent and properly integrated."
Section 3.4.4, "Interface Control," states that interface procedures between RG&E engineering and contractor engineering organizations shall include "instructions regarding the contents of document transmittals with consideration for response requirements.
Transmittals of design documents shall identify the status of the documents and identify, where necessary, incomplete items which require further evaluation, review, or approval."
Contrary to the above, the licensee was not properly controlling, verifying, and accepting design reports, calculations, or analyses.
Specifically, (1)
The NUS Corporation calculation supporting Engineering Work Request (EWR) 1594, "Hydraulic Analysis of the Service Water System," dated February 1988, was submitted to the licensee marked "preliminary, for review and comment".in March 1988.
As of December 1991, the analysis had not been reviewed and accepted by the licensee, but was being used for hydraulic analysis and balancing of the service water system (SWS) following installation of spent fuel pool heat exchanger B.
(2)
Bechtel-KWU Report, "Heat Load Capacity/Design Margin Analysis for RHRHX, CCWHX, SFPHX, Non-Regenerative HX, "Job No. 20031, Rev. 1, dated January 22, 1989, did not have any indication that it had been reviewed and accepted.
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Enclosure 1
(3)
Two copies of the RGB'esign analysis for EWR 3689, "Containment Fan Cooler AirFlow," Rev. 0, dated June 4, 1984, were in existence.
One contained handwritten notes and corrections, the other did not.
The licensee stated that the correct design analysis was the one with handwritten additions even though they had not been reviewed and approved as part of a new revision.
(4)
Rochester Gas and Electric Corporation (RG&E) design analysis document entitled "Insitu Motor Load Determinations" for EWR 4232, "SWS Pump Motor Studies," dated July 15, 1986, included incorrect assumptions based on a low slip motor and inappropriate equations.
Consequently, the results were also incorrect.
The incorrect assumptions had not been identified by the design verification process.
(5)
RGB'esign analysis for EWR 4658-ME-009, "Minimum Diesel Generator Cooler and Lube Oil Cooler Water Flow Requirements,"
Rev. 0, dated July 25, 1991, was prepared to justify DG operability with high differential pressure (dp) across the coolers due to zebra mussel tube plugging.
The calculations did not use the applicable design conditions for cooling water inlet temperature and heat load.
The incorrect design values resulted in incorrect assumptions and results.
This is a Severity Level IVviolation. (Supplement Il)
B.
10 CFR 50.34(b) states, in part, "The final safety analysis report shall include information that describes the facility, presents the design bases and the limits on its operation, and presents a safety analysis of the structures, systems and components.
~.and shall include...a description and analysis of the...components of the facility, with emphasis upon performance requirements, the bases, with technical justification therefore upon which such requirements have been established, and the evaluations required to show that safety functions willbe accomplished.
The description shall be sufficient to permit understanding of the system designs and their relationship to safety evaluations."
10 CFR Part 50.71(e) states, in part, "Each person licensed to operate a nuclear power reactor...shall update periodically, as provided in paragraphs (e) (3) and (4) of this section, the final safety analysis report (FSAR) originally submitted as part of the application for operating license, to assure that the information included in the FSAR contains the latest material developed...The updated FSAR shall be revised to include the effects of all changes made in the facility or procedures as described in the FSAR...revisions shall be filed no less frequently than annually and shall reflect all changes up to maximum of 6 months prior to the date of filing..."
Enclosure 1
Contrary to the above, a number of discrepancies existed in the Ginna Updated Final Safety Analysis Report discussion of the SWS.
(1)
USFAR Section 9.2.1.3 stated "...the service water loop is isolated by normally closed valves to provide two independent systems with no sizable cross-connections."
The SWS has been cross connected by the 14" supply header for the containment air coolers since approximately March 1988.
The system is also cross-connected at the three inch equipment cooler supply
- headers, at the three inch SI pump supply headers, and at the four inch component cooling water cross-connect in the supply header.
(2)
USFAR Section 9.2.1.3 stated"...All engineered safety features equipment is split between the two systems so that only half of the system would be affected by a malfunction."
Due to the system cross connects, the equipment is not split between two systems.
Also, ifthe system were operated split, the three safety injection pumps would all be on one header since they cannot be divided between headers.
(3)
USFAR Sections 9.2.1.2.1 and 9.2.1.2.2 stated that the service water system is designed to isolate non-safety-related loads on an accident and a safety injection signal, respectively.
The SWS does not isolate non-safety-related loads on a safety injection signal (SIS) unless an undervoltage condition also exists.
(4)
UFSAR Table 9.2-2 did not accurately reflect the total flow to the containment air coolers (CACs) because it neglected the flow to the CAC fan motors.
This is a Severity Level IV violation.
(Supplement VII)
C.
10 CFR 50, Appendix B, Criterion III, "Design Control," requires that design control measures"...provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program,"
10 CFR 50, Appendix B, Criterion XI, "Test Control," requires, in part, that test results be documented and evaluated to ensure satisfactory completion of test requirements.
Contrary to the above, the licensee had not reviewed the preoperational test results in comparison to current system operation and configuration and determined the need for additional testing to support operation in required system configurations.
This had not been done despite system operating changes since initial licensing involving the number of pumps normally operating (changed from three to two) and various changes to the system valve alignment.
Specific test deficiencies include:
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Enclosure 1
(1)
The non-safety-related loads were isolated from the loop headers when the safety-related performance of the system was tested (single pump operation).
Current operation of the plant does not isolate the non-safety-related loads during an SIS, (2)
The safety-related performance of the system was not tested with two pumps operating.
Two pumps are required to handle the post-accident heat load during recirculation.
(3)
The system flow balance was established based on three pump operation, not based on the limiting case of one pump operating supplying all safety-related and non-safety-related loads.
(4)
Pump run-out conditions were not evaluated or considered.
This is a Severity Level IV violation.
(Supplement II) 4 Pursuant to 10 CFR 2.201, Rochester Gas and Electric is required to submit to this office within 30 days of the date of the letter transmitting this Notice, a written reply, including:
(1) the reasons for the violations, ifcontested the basis for disputing them, (2) the corrective steps that have been taken and the results achieved, (3) the corrective steps to be taken to avoid further violations, and (4) the date when full compliance willbe achieved. Ifgood cause is shown, consideration willbe given to extending the response time.
The reply directed above is not subject to clearance by the Office of Management and Budget under the Paperwork Reduction Act of 1980, Pub. L.96-511.