View source for BSEP-88-1234, Responds to NRC Re Violations Noted in Insp Repts 50-324/88-34 & 50-325/88-34.Corrective Action:Review Conducted of Administrative & Engineering Procedures in Place at Time of Initiation & Approval

{{Adams
| number = ML20196A975
| issue date = 11/30/1988
| title = Responds to NRC Re Violations Noted in Insp Repts 50-324/88-34 & 50-325/88-34.Corrective Action:Review Conducted of Administrative & Engineering Procedures in Place at Time of Initiation & Approval
| author name = Harness J
| author affiliation = CAROLINA POWER &amp; LIGHT CO.
| addressee name = 
| addressee affiliation = NRC OFFICE OF ADMINISTRATION &amp; RESOURCES MANAGEMENT (ARM)
| docket = 05000324, 05000325
| license number = 
| contact person = 
| document report number = BSEP-88-1234, NUDOCS 8812060157
| title reference date = 11-30-1988
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE, UTILITY TO NRC
| page count = 4
}}

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Cp&L Carolina Power & Light Company Brunswick Steam Electric Plant P. O. Box 10429 Southport, NC 28461-0429 November 30, 1938 FILE:   B09-13510C                                                   10CFR2.201 SERIAL:   BSEP/68 1234 U.S. Nuclear Regulatory Commission ArrN: Document Control Desk Washington, DC 20555 BRUNSVICK STEAM ELECTRIC PLAh7 UNITS 1 AND 2 DOCKET NOS. 50-325 AND 50-324 LICENSE NOS. DPR-71 AND DPR-62 RESPONSE TO INyRACTIONS OF NRC REQUIREMENTS Gentlemen:
The Brunswick Steam Electric Plant (BSEP) has received I&E Inspection Report 50-325/88-034 and 50-324/88-034 and finds that it does not contain information of a proprietsry nature.
This report idsntified one item that appeared to be in noncompliance with NRC requirements. Enclosed is Carolina Power & Light Company's response to this violation.
Very truly yours.
L. Harness, General Manager Brunswick Steam Electric Plant TH/ meg Enclosure cc:   Mr. B. C. Buckley Mr. M. L. Ernst BSEP NRC Resident Office SS12060157 DR ADocg 881130
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VICLATION 10CFR50, Appendix B, Criterion III, Design Control, requires that effective measures be established to assure that applicable regulatory requirements and design bases are correctly translated into drawings.                                                        Such design bases included that of Technical Specification 3.6.1.2, which states that the calculated peak containment internal pressure related to the design basis accident, Pa, is 49 psig.
Contrary to the above, effective measures to incorporate design bases into drawings were not established in that drawing FP-9700 for primary containment boundary valve 1/2 E51-F040 (reactor core isolation caoling steam exhaust line check valve) specified a 25 psig disk design maximum pressure instead of Pa.
l This is a Severity Level IV violation (Supplement I).

===RESPONSE===
l l                                  I. Admission or Denial of the Alleged Violation l
l                                         CP&L acknowledges the violation of not having established effective measures to incorporate design bases into drawings concerning drawing FP-9700 as described.
II. Reason for the Violation Plant Modifications 81-274 (Unit 2) and 81-275 (Unit 1) replaced the E51-F040 reactor core isolation cooling (RCIC) turbine exhaust line swing check valve with the lif t-check design. The valves had experienced disc slamming and consequent damage resulting from RCIC operation. The design basis document for the modification addressed operational exhaust pressures under various turbine steam supply pressures but failed to address containment accident pressure of LLRT test pressure for the specified valve. Neither the mod engineer nor the ISI engineer during their reviews of the modification addressed the discrepancy between the test pressure requirements of the LLRT procedure, PT-20.3.160 (which requires e test pressure of 58 psig), and the stated design pressure of the disc of E51-F040 on drawing FP-9700 (which states that the design pressure for the disc in 25 psig).
In accordance with applicable codes and specifications, the valve body, as supplied, was shop hydro tested to 450 psi and received a postinstallation acceptance hydro test of 225 psi. The valve disc was hydro tested to 38 psi, as supplied, and was local leak rate tested after installation at 58 psi (49 psig actual differential pressure on the disc) per the modification acceptance test. At no time during either the mod reviews or the acceptance testing was the discrepancy identified.

s III. Corrective Steps Which Have Been Taken and Results Achieved Following identification of the discrepancy, NCR A-88-037 was written to document the concern. EER 88-0461 was generated to address and resolve the concerns associated with the drawing discrepancies. The valve manufacturer, Anchor-Darling, was contacted about the concern. Their local leak rate responseAindicated testing.             thatresponse subsequent   the disc clarified would not that   be damaged the discby"will withstand a                                   i pressure of 62 psig at 248 degrees F for a sustained period of time."                                                  l These values are greater than the containment system requirements for the design basis accident as defined in the FSAR and higher than credible turbine exhaust operating. Based on the review of LLRT testing data                                         and the disc evaluation by the valve manufacturer, the presently installed valve haa been evaluated as acceptable. No change in spare part number for the disc will be required as the present valve disc is suf ficient for                                            ;
an application requiring a design pressure of 62 psig.           Plant Drawine Change (PDC) 88-0123 added a note to FP-9700 stating that the disc will withstand 62 psig at 248 degrees F.                                                                                    i To ensure that a similar situation does not exist on the comparable high pressure coolant injection (HPCI) turbine exhaust line check valves, PMs 86-039 and 87-199, which replaced the HPCI swing-check valves with the lift-check design, were reviewed. Stated design values for the lift-check disc on FP-82383, Rev. A (9/28/87), encompass LLRT test pressure and containment system requirements.
IV. Corrective Action Which Will be Taken to Avoid Further Violations and Date When Full Compliance Vill be Achieved A review was conducted of the administrative and engineering procedures in place at the time of PMs 81 274 and 81-275 initiation and approval.                                                l Guidelines for design and design review included specific reference to                                                l ASSI-45.2.11 which includes a checklist of technical considerations to be invoked during the design and design review process of a plant                                                        [
modification. Training instruction TI 601 was also in effect, which                                                   t required a signature by the system engineer that he/she had become                                                    i f amiliar with the FSAR and Technical Specifications, which include the design requirements of containment. Furthermore, the plant modification received a review by an ISI engineer as indicated on the plant modifica-                                              i tion traveler.                                                                                                        l l
While administrative procedures in place at the time of the modification                                              i should have been sufficient to prevent the installation of a disc with a                                              l stated design pressure less than that required by the application, subsequent revisions to applicable administrative and engineering                                                     ;
procedures have added formalized checklists and training requirements for safety reviewers. ENP-03 now requires a signature by both a first and 1
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  ..                 g second safety reviewer for Q-list components, which includes containment isolation valves (a safety reviewer's qualification is determined by the training requirements of TI-302). ENP-03, Revision 33, formalized the ANSI-45.2.11 checklist which is to be utilized during the design and technical review process and includes specific codes and off-normal conditions that a component, system, or structure may be subjected to.
The safety reviewer qualification process has additionally been revised to include retraining and requalification as described in TI-302 Revision 1, approved July 21, 1987.
To ensure that the guideline checklist Figure 1 of ENP-03 is understood by engineers involved in the preparation and review of plant modifica-tions, real-time training will be conducted covering the design input considerations of ANSI-45.2.11 and Figurs 1 of the current revision of ENP-03 (Rev. 36). The NCR for this incident (NCR A-88-037) will also be reviewed with Technical Support personnel in real-time training. Training will be completed by December 30, 1988.
Additionally, the BESU and NED modification groups will review this incident (NCR A-88-037) through their required reading programs, BESU reviews were completed on November 29, 1988, and NED reviews will be completed by December 30, 1988.
Based on the requirements of training indicated for plant modification preparers and reviewers, CP&L will be in full compliance by December 30, 1988.
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