ML20148E233

Notice of Violation from Insp on 970310-27.Violation Noted: Licensee Had Not Updated FSAR to Assure That Info Included Contains Latest Matl Developed

ML20148E233
Person / Time
Site:	Grand Gulf
Issue date:	05/29/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20148E230	List: ML20148E226 ML20148E233 ML20148E244
References
50-416-97-05, 50-416-97-5, NUDOCS 9706030037
Download: ML20148E233 (5)

Text

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ENCLOSURE 1 NOTICE OF VIOLATION Entergy Operations, Inc. Docket No.: 50-416 Grand Gulf Nuclear Station License No.: NPF-29 i During an NRC inspection conducted on March 10-27,1997, three violations of NRC requirements were identified. In accordance with the " General Statement of Policy and 4

Procedure for NRC Enfurcement Actions," NUREG-1600, the violations are listed below:

A. 10 CFR Part 50, Appendix B, Criterion Ill, requires that measures shall be established to assure that applicable regulatory requirements and the design basis j for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. The design control measures shall also provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or l simplified calculational methods, or by the performance of a suitable testing program. Design changes, including field changes, shall be subject to design control measures commensurate with those applied to the original design.

Contrary to the above, as of March 27,1997, the following design control deficiencies were identified:

1. The design control measures did not assure that the low discharge pressure alarm setpoint in Alarm Response Procedure 04-1-02-1 H13-P870-1 A-D1, "SSW Pmp A Disch Pres Lo," Revision 100, and Alarm Response Procedure 04-1-02-1 H13-P870-7A-D1, "SSW Pmp B Disch Pres Lo,"

Revision 100, was updated when the standby service water system pressure was raised by Design Change Package 82/5020, which replaced the original service water pumps with higher performance pumps prior to initiallicensing.

2. The design control measures did not assure that the high leakage alarm setpoi.nt in Alarm Response Procedure 04-1-02-1 H13-P870-1 A-E1, "SSW Loop A Leak Hi," Revision 100, and Alarm Response Procedure 04-1-02-1 H13-P870-7A-E1, "SSW Loop B Leak Hi," Revision 100, was updated when the setpoint was raised from 950 gpm to 1200 gpm by Design Change Package 81/5015, which was related to the service water pump replacement.

3. The design control measures did not assure that Calculation MC-Q1P41-86007, " Standby Service Water Ultimate Heat Sink Performance," Revision 0, July 25,1986, considered the worst-case design basis conditions as required by General Design Criterion 44.

Specifically, the calculation did not conservatively identify the heat rejection rate for the ultimate heat sink, assuming that both trains of onsite power were available.

9706030037 970529 PDR ADOCK 05000416 G PDR

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l 4. The design control measures did not assure that the design basis requirement l to prevent water from reaching the floor-mounted, safety-related equipment in the standby service water pump house, as discussed in Updated Final l Safety Analysis Report Section 2.4.10, " Flooding Protection Requirements,"

l was adequately implemented in the current design. Specifically, the current ,

design did not include a drain path for normal and expected pump shaft seal l leakage.

l This is a Severity Level IV violation (Supplement 1)(50-416/9705-01).

B. 10 CFR 50.71(e) requires that the licensee periodically update the Final Safety Analysis Report, orininally submitted as part of the application for an operating license, to assure t. v. the information included in the Final Safety Analysis Report contains the latest material developed.

, Contrary to the above, as of March 27,1997, the licensee had not updated the l Final Safety Analysis Report to assure that the information included contains the latest material developed. For example:

1. The licensee failed to update the safety analysis report flood analysis for the

! service water pumphouse when the site drainage was improved and the potential for flooding was reanalyzed.

2. The licensee f ailed to ensure that Table 9.2-2, " Standby Service Water System Active Failure Analysis," contained the latest material developed.

Specifically, the next to last item incorrectly stated that the redundant cooling tower will meet the cooling requirements for a cooling tower fan failure. This statement is not correct because if a fan fails on Basin A (which also serves the high pressure core spray service water system) and no operator action is taken to reduce the heat load on the Standby Service

, Water Loop A, the basin temperature may exceed the design temperature of 90 F due to the reduction in evaporative cooling with the Standby Service Water Loop A operating only with natural convection cooling.

3. The licensee f ailed to ensure that Table 9.2-3, " Standby Service Water System Loads for Nominal Flow Conditions," correctly listed the loads on each of the heat exchangers served by the standby service water system.

Specifically, the table indicates that the design heat load for the fuel pool heat exchangers was 7.45 million BTU /hr at 30 days after shutdown. ,

However, Table 9.2.17, " Time Dependent Standby Service Water System Cooling Duty Loads Following a DBA (LOCA With a Loss of Offsite Power l and Single Active Failure)," indicates that this value was i 13.13 million BTU /hr. Further, the licensee's high density fuel licensing

! submittal of November 1,1991, described this load as 13.39 million BTU /hr l at 37 days after shutdown.

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4. The licensee f ailed to ensure that Table 9.2-16, " Standby Service Water System Cooling Duty Loads Following a DBA (LOCA With Loss of Offsite Power and Failure of Standby Diesel Generator A)," correctly listed the fuel pool heat load. Specifically, the table listed the heat load as 7.45 million BTU /hr and the associated note stated that the load as 13.13 million BTU /hr based on 18 core off-loads. However, the licensee indicates that with the extended fuel cycle only 12 of the larger offloads would fill the pool, and that this was the correct basis for the l 13.13 million BTU /hr value. As stated in example 3, the correct value for l the current licensing basis heat load at 37 days after shutdown was actually I 13.39 million BTU /hr. Therefore, both the heat load and the attendant fuel cycle basis were incorrect.

5. The licensee f ailed to ensure that Table 9.2-17, " Time Dependent Standby Service Water System Cooling Duty Loads Following a DBA (LOCA With a Loss of Offsite Power and Single Active Failure)," listed the correct fuel pool heat load. Specifically, the table listed the fuel pool heat load as 13.13 million BTU /hr and the associated note incorrectly stated that it was l based on 18 core off-loads. As described above, the current Icensing basis heat load was 13.39 million BTU /hr at 37 days after shutdown and the basis was 12 of the extended cycle offloads.

6. The licensee failed to ensure that Figures 9.2-5, " Total Heat Rejection Rate, Basin A (Unit 2 Loop A and Unit 1 Loop C), Time Period 0 to 30 Days," and 9.2-6, " Total Heat Rejection Rate, Loop B (Unit 1 Loop B and Unit 2 Loop B),

Time Period - O to 30 Days," was updated, in that they showed 2-unit heat loads and the Grand Gulf Nuclear Station was a single-unit plant.

7. The licensee failed to ensure that Figure 9.2-6a, " Total Heat Rejection Rate, Basin B, Loop B and Loop C, Time Period - O to 30 Days," showed the correct peak heat rate. Specifically, the table listed the peak heat rate as approximately 27 million BTU /hr (74 thousand BTU /sec); however, Figure 6 of Calculation MC-Q1P41-86054, " Standby Service Water Ultimate Heat Sink Performance," Revision 0, dated July 25,1986, listed the peak heat load as approximately 23 million BTU /hr. The Updated Final Safety Analysis Report figure appeared to be incorrect.

This is a Severity Level IV violation (Supplement I) (50-416/9705-03).

. C. 10 CFR Part 50, Appendix B, Criterion XI, " Test Control," required that "A test l program shall be established to assure that all testing required to demonstrate that I structures, systems, and components will perform satisf actorily inservice is j identified and performed per test procedures which incorporate the requirements I and acceptance limits contained in the applicable design documents." It further l required that, " Test procedures shall include provisions for assuring

! that.. adequate test instrumentation is available and used."

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i l l Contrary to the above, as of March 27,1997, the test control program did not establish acceptance limits contained in applicable design documents and the test control program did not include provisions for assuring that adequate test instrumentation is available and used. For example:

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1. Surveillance Procedures 06-OP-1P41-0-0004, " Standby Service Water Loop A Valve and Pump Operability Test," Revision 102, dated August 16, 1996, and 06-OP-1P41-0-0006, "HPCS Service Water System Valve and

! Pump Operability Test," Revision 102, dated December 14,1996, did not

( inclur'e acceptance limits which would assure pump performance at the l levres assumed in the accident analysis, in addition, the acceptance limits di.f not include a.i adjustment to compensate for instrument errors.

2. Performance and System Engineering Procedures 17-S-06-22, "SSW "A" Performance," Revision 4, dated April 19,1995, and 17-S-06-23, "SSW "B" Performance," Revision 4, dated September 28,1995, did not provide i adequate documentation that the results were evaluated to assure that test l requirements have been satisfied. I l

3. Engineering Standard GGNS-MS-39.0, "Machanical Standard for Thermal Performance Testing of Safety-Related Standby Service Water Heat Exchangers," Revision 0, dated February 26,1992, did not use the current design basis requirements as the basis for the acceptance limits for the fuel pool cooling and c!cenup heat exchanger.

4. Engineering Standard GGNS-MS-39.0, " Mechanical Stancard for Thermal Performance Testing of Safety-Related Standby Service Water Heat Exchangers," Revision 0, dated February 26,1992, had not been updated to reflect changes in instrumentation practices related to measuring standby service water inlet temperature or the use of capillary tube thermometers. In addition, the heat transfer values in the standard did not include an adjustment to compensate for instrument errors.

This is a Severity Level IV violation (Supplement 1)(50-416/9705-06).

Pursuant to the provisions of 10 CFR 2.201, Entergy Operations, Inc., is hereby required to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555 with a copy to the Regional Administrator, Region IV,611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011, and a l copy to the NRC Resident Inspector at the f acility that is the subject of this Notice, within 30 days of the date of the letter transmitting this Notice of Violation (Notice). 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 i

, compliance will be achieved, in addition, for Violation C, Example 3, your response should l

4 generically address the failure to establish acceptance limits consistent with design basis requirements. 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 time specified in this Notice, an order or a Demand for Information i

may be issued as to wh/ 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.

! Because your response will be placed in the NRC Public Document Room (PDR), to the 2

extent possible, it should not include any personal privacy, proprietary, or safeguards

information so that it can be placed in the PDR without redaction. However, if you find it necessary to include such information, you should clearly indicate the specific information that you desire not to be placed in the PDR, and provide the legal basis to support your i request for withholding the information from the public.

i Dated at Arlington, Texas this)@ day of 1997