NRC Generic Letter 1986-07

NRC Generic Letter 1986-007: Transmittal of NUREG-1190 Regarding San Onofre Unit 1 Loss of Power & Water Hammer Event

ML031150288
Person / Time
Site:	Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, 05000000, Zimmer, Fort Saint Vrain, Washington Public Power Supply System, Shoreham, Satsop, Trojan, Clinch River
Issue date:	03/20/1986
From:	Harold Denton Office of Nuclear Reactor Regulation
To:
References
NUREG-1190 GL-86-007, NUDOCS 8603210334
Download: ML031150288 (7)

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D. C. 20555 A March 20, 1986 TO ALL REACTOR LICENSEES AND APPLICANTS

Gentlemen:

SUBJECT: TRANSMITTAL OF NUREG-1190 REGARDING THE SAN ONOFRE UNIT 1 LOSS OF POWER AND WATER HAMMER EVENT (GENERIC LETTER 86-07)

On November 21, 1985, while operating at 60% power, Southern California Edison Company's San Onofre Unit 1 Nuclear Power Plant experienced a loss of ac electrical power followed by a severe water hammer in the secondary system which caused a steam leak and damaged plant equipment. Shortly after the event, the NRC Executive Director for Operations directed that an NRC Team be sent to San Onofre, in conformance with the recently established Incident Investigation Program, to investigate the circumstances of this event. The NRC Team has now completed its investigation and has documented the factual information and their findings and conclusions associated with the event (see enclosed NUREG-1190,

entitled "Loss of Power and Water Hammer Event at San Onofre Unit 1, on, November 21, 1985").

In this report, the team has concluded that the event was significant because (a) all inplant ac power was lost for 4 minutes; (b) all steam generator feedwater was lost for 3 minutes; (c) a severe water hammer caused by check valve failures was experienced in the feedwater system which caused a leak, damaged plant equipment and challenged the integrity of the auxiliary feedwater system;

(d) all indicated steam generator water levels dropped below scale; and (e)

the reactor coolant system experienced an acceptable but unnecessary cooldown transient. In the team's view the most significant aspect of the event was that five safety-related feedwater system check valves degraded to the point of inoperability during a period of less than a year, without detection, and that their failure jeopardized the integrity of safety-related feedwater piping.

The cause of the feedwater system check valve failures has been preliminarily identified by SCE as partial or complete separation of the check valve disc assemblies due to fluid flow conditions. Information submitted to the staff on this subject is currently under review.

You should review the information in the enclosed report for applicability to your facility. In addition, you should ensure that the information in NUREG-1190

is made available to your plant staff as part of your training program in connection with the Feedback of Operating Experience to Plant Staff (TMI Action Plan Item I.C.5).

8@<)ijvf0 Wh&7/t?

P>A~c >OSDOOOoOf

March 20, 1986

-2- On February 4, 1986, the Executive Director for Operations (EDO) identified and assigned responsibility for generic and plant-specific actions resulting from the investigation of the San Onofre event. Some of the generic actions may be applicable to your facility. A copy of the EDO memorandum is included for your information.

This generic letter is provided for information only, and does not involve any reporting requirements. Therefore, no clearance from the Office of Management and Budget is required. The enclosed report is currently under NRC review.

Any generic requirements stemming from the report will be transmitted at a later date following completion of the appropriate procedural steps.

Sincerely, Harold R. Denton, Director

-Offic of Nuclear Reactor Regulation Enclosures:e>l6,O> ( ) A

1. NUREG-1190

2. EDO Memorandum of February 4, 1986

3. List of Generic Letters

STAFF ACTIONS RESULTING FROM THE INVESTIGATION

OF THE NOVEMBER 21 SONGS-1 EVENT

(Reference: NUREG-1190)

1. Item: Adequacy of feedwater check valves to perform safety function.

(References: Commission briefing, Sections 6.2.4, 6.4, 6.7, and Principal Finding)

Action Responsible Office Category Implement and coordinate IE Plant-specific the staff and industry actions Generic necessary to assure the reliability of safety-related check valves.

Other offices to assist as requested. Areas to be evaluated include:

- licensee's engineering report on root cause analysis and proposed corrective actions

- adequacy of check valve design for this applica- tion

- adequacy of Inservice Testing (IST) Program and procedures to detect degraded and failed valves

- adequacy of check valves (and related testing programs) in other systems such as RHR system

2. Item: Completeness of resolved USI A-1, "Water Hammer".

(References: Finding numbers 1, 2, 3, 8 and 9)

Action Responsiblf *Office Category Assess the need to re- NRI Generic evaluate USI A-1 to specifically address the potential for and prevention of condensation-induced water hammers in feedwater piping (assume the issue concerning check valve integrity will be resolved in item 1).

- 2 -

3. Item: Adequacy of San Onofre Unit 1 design.

(Commission briefing, Finding numbers 11 and 13)

Action Responsible Office Category Implemert and coordinate NRR Plant-specific the staff's actions to re-evaluate the following San Onofre design features:

manual loading of the diesel generators follow- ing a loss of power event manual actuation of steam line isolation valves and assurance of steam generator availability to remove decay heat lack of steam generator blowdown status in control room adequacy of the licensee's design change to eliminate spurious SI indication on loss of ower

4. Item: Ade'cy of post-trip review.

(References: Sections 6.6 and 7.2.2.4 and Finding number 17)

Action Responsible Office Category- a. Evaluate NRC require- NRR Generic ments for ensuring that sufficient event data are retrievable to accurately reconstruct the event following a loss of offsite power.

b. Evaluate the licensee's Region V Plant-specific process for post-trip review and evaluation, including the thoroughness of review and oversight provided by the onsite and offsite nuclear safety review groups.

_ 3 -

5. Item: Adequacy of licensee's recordkeeping practices.

(References: Section 6.5 and Finding number 20)

Action Responsible Office Category Evaluate the adequacy of Region V Plant-specific the licensee's maintenance records.

6. Item: Adequacy of operator training and/or procedures.

(References: Section 7 and Finding numbers 14, 15 and 16)

Action Responsible Office Category Review the implementation Region V Plant-specific of the training program regarding operator under- standing and actions in the area of electrical systems, and invoking technical specification action statements.

7. Item: Adequacy of emergency notifications and NRC response.

(References: Section 7.3 and Finding number 22)

Action Responsible Office Category a. Verify the adequacy of Region V Plant-specific the licensee's procedures and training for reporting of events to NRC Operations Center.

b. Evaluate the need for IE Generic changes in NRC policy or guidance regarding: the use of the ENS line; the use of NRC personnel as ENS communicators; and possible approaches to improve the ability to determine the overall plant status.

-4 -

8. Item: Significance of backlog of license amendments.

(Reference: Commission briefing)

Action Responsible Office Ca tegoty Evaluate whether a backlog NRR

of license amendments and Plant-specific technical specification changes contributed to delays in approving the licensee's.

IST program.

List of Recently Issued Generic Letters Generic Letter No. Subject Date of Issue Issued To

86-06 To be Issued

86-05 To be Issued

86-04 Policy Statement on 02/13/86 All Power Reactor Engineering Expertise Licensees and on Shift Applicants for Power Reactor Licenses

86-03 Applications for License 02/10/86 All Power Reactor Amendments Licensees and OL

Applicants

86-02 Technical Resolution of 01/23/86 All Licensees of Generic Issue B-19 - Operating BWRs Thermal Hydraulic Stability

86-01 Safety Concerns Associated 01/03/86 All BWR Applicants with Pipe Breaks in the and Licensees BWR Scram System

85-22 Potential for Loss of Post - 12/03/85 All Licensees of LOCA Recirculation Capability Operating Reactors, Due to Insulation Debris Applicants for OLs Blockage and Holders of CPs

85-21 Not Issued

85-20 Resolution of Generic Issue 10/30/85 All Licensees with

69: High Pressure Injection/ Babcock and Wilcox Make-up Nozzle Cracking in Operating Reactors Babcock and Wilcox Plants