Text

Conformance to NRR Generic Ltr 82-16,HB Robinson Steam Electric Plant,Unit 2, Technical Evaluation Rept
	ML14183A025
Person / Time
Site:	Robinson
Issue date:	12/31/1983
From:	Beahm D; IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:	; NRC
Shared Package
ML14183A023	List: ML14183A024; ML14183A025;
References
	CON-FIN-A-6600, RTR-NUREG-0737, RTR-NUREG-737 EGG-EA-6437, GL-82-16, NUDOCS 8510040100
	Download: ML14183A025 (21)
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Enclosure EGG-EA-6437 DECEMBER 1983 CONFORMANCE TO NRR GENERIC LETTER 82-16.

H. B. ROBINSON STEAM ELECTRIC PLANT UNIT 2 D. M. Beahm Idaho National Engineering Laboratory Operated by the U.S. Department of Energy Prepared for the~

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This is an informal report intended for use as a preliminary'or working document,

.85104100 850912 PDR ADOCdK.05000261.1 P pP-DR Prepared for the 1U. S. NUCLEAR REGULATORY COMMISSION f

Under DOE Contract No. DE-AC07-761DO1570

,ECQdh FIN No. A6600 E

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CONFORMANCE TO NRR GENERIC LETTER 82-16 H. B. ROBINSON STEAM ELECTRIC PLANT UNIT 2 David M. Beahm Published December 1983 EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Atlanta, Georgia 30303 Under DOE Contract No. DE-ACO7-761D01570 FIN No. A6600

ABSTRACT This EG&G Idaho, Inc., report evaluates the submittal provided by Carolina Power and Light Company (CP&L) for H. B. Robinson Steam Electric Plant Unit 2 (HBR-2). The submittal is in response to Generic Letter No. 82-16, "NUREG-0737 Technical Specifications (TS)."

Applicable sections of the plant's TS are evaluated to determine compliance to the guidelines established in the generic letter.

FOREWORD This report is supplied as part of the "Technical Assistance for Operation Reactors Licensing Actions" being conducted for the U.S. Nuclear Regulatory Commission Region II by EG&G Idaho, Inc., NRC Licensing Support Section.

The U.S. Nuclear Regulatory Commission Funded the work under authorization B&R 92-19-20-10, FIN No. A6600.

Docket No. 50-261 TAC No. 49753 li

CONTENTS ABSTRACT 11 FOREWORD 11

1. INTRODUCTION..................................................

2. REVIEW REQUIREMENTS............................................

2 2.1 STA Training (I.A.1.1.3).................................. 2 2.2 Shift Manning-Overtime Limits (I.A.1.3.1)..................

2 2.3 Short Term Auxiliary Feedwater System (AFWS) Evaluation (II.E.1.1)..............................................3 2.4 Safety Grade AFW Initiation and Flow Indication (II.E.1.2)..............................................3 2.5 Dedicated Hydrogen Penetrations (II.E.4.1).................

3 2.6 Containment Pressure Setpoint (II.E.4.2.5).................

3 2.7 Containment Purge Valves (II.E.4.2.6)...................... 4 2.8 Radiation Signal on Purge Valves (II.E.4.2.7)..............

4 2.9 Upgrade Babcock and Wilcox (B&W) AFWS (II.K.2.8)...........

4 2.10 B&W Safety-Grade Anticipatory Reactor Trip (II.K.2.10).....

5 2.11 B&W Thermal-Mechanical Report (II.K.2.13)..................

5 2.12 Reporting Safety and Relief Valve Failures and Challenges (II.K.3.3)..............................

5 2.13 Anticipatory Trip on Turbine Trip (II.K.3.12)..............

5

3. EVALUATION...................................................7 3.1 STA Training (I.A.1.1.3).................................. 7 3.2 Shift Manning-Overtime Limits (I.A.1.3.1)..................7 3.3 Short Term Auxiliary Feedwater System (AFWS) Evaluation (II.E.1.1) 8 3.4 Safety Grade AFW Initiation and Flow Indication (II.E.1.2)..............................................

3.5 Dedicated Hydrogen Penetrations (II.E.4.1).................

8 3.6 Containment Pressure Setpoint (II.E.4.2.5).................

9 3.7 Containment Purge Valves (II.E.4.2.6)......................

9 3.8 Radiation Signal on Purge Valves (II.E.4.2.7)..............

10 3.9 Upgrade Babcock and Wilcox (B&W) AFWS (II.K.2.8)...........

10 3.10 B&W Safety-Grade Anticipatory Reactor Trip (II.K.2.10).....

10 3.11 B&W Thermal-Mechanical Report (II.K.2.13).................. 10 3.12 Reporting Safety and Relief Valve Failures and Challenges (II.K.3.3).................................... 11 3.13 Anticipatory Trip on Turbine Trip (II.K.3.12)..............

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4. CONCLUSIONS...........................

....................... 13

5. REFEkENCES................................................... 15 iv

CONFORMANCE TO NRR GENERIC LETTER 82-16 H. B. ROBINSON STEAM ELECTRIC PLANT UNIT 2

1. INTRODUCTION On September 20, 1982, Generic Letter 82-161 was issued by D. G. Eisenhut, Director of Licensing, Office of Nuclear Reactor Regulation (NRR), to all pressurized power reactor licensees. This letter identified a number of items which were required by NUREG-07372 to be implemented into the licensee's Technical Specifications (TS) by December 31, 1981.

Each licensee was requested to review his facility's TS, to address areas of compliance, and to identify deviations or absence of a specification for the items identified in the generic letter, within 90 days of receipt of the letter.

The Carolina Power and Light Co (CP&L), the licensee for H. B. Robinson Steam Electric Plant Unit 2 (HBR-2), provided a response to the generic letter on December 23, 1982.3 This report provides an evaluation of the licensee's TS and Nuclear Regulatory Commission (NRC) correspondence with the licensee pertaining to those items identified in the generic letter.

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2. REVIEW REQUIREMENTS The review consists of evaluating the licensee's response, currently approved TS, and other NRR approvals against the criteria set forth in Generic Letter 82-16. The NUREG-0737 items and the criteria established are as follows:

2.1 STA Training (I.A.1.1.3)

The licensee is to address within his TS that a shift technical advisor (STA) to the shift supervisor is provided. In addition, the qualifications, training, and on-duty requirements for the STA should be stated.

2.2 Shift Manning-Overtime Limits (I.A.1.3.1)

The licensee is to provide changes to his TS providing overtime administrative procedure and staffing requirements. The following guidelines were established for the licensee by the NRC.

"a. An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight (excluding shift turnover time).

b. An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24-hour period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48-hour period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any seven day period (all excluding shift turnover time).

c. A break-of at least eight hours should be allowed between work periods (including shift turnover time).

d. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.

Recognizing that very unusual circumstances may arise requiring deviation from the above guidelines, such deviation shall be authorized by the plant manager or his deputy, or higher levels of management. The paramount consideration in such authorization shall be that significant reductions in the effectiveness of operating personnel would be highly unlikely.

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In addition, procedures are encouraged that would allow licensed operators at the controls to be periodically relieved and assigned to other duties away from the control board during their tour of duty." 4 2.3 Short Term-Auxiliary Feedwater System (AFWS) Evaluation (II.E.1.1)

The objective of this item is to improve the reliability and performance of the auxiliary feedwater (AFW) system. TS depend on the results of the licensee's evaluation and the staff review, and are being developed separately for each plant. The limiting conditions of operation (LCO's) and surveillance requirements for the AFW system should be similar to other safety-related systems.

2.4 Safety Grade AFW Initiation and Flow Indication (II.E.1.2)

The AFW system automatic initiation system was to have been control grade by June 1, 1980, and safety grade by July 1, 1981; the AFW system flow indication was to have been control grade by January 1, 1980, and safety grade by July 1, 1981.1 2.5 Dedicated Hydrogen Penetrations (II.E.4.1)

Plants that use external recombiners or purge systems for post-accident combustible gas control of the containment atmosphere should provide containment penetrations dedicated to that service. In satisfying this item, some plants may have to add some additional piping and valves.

If so, these valves should be subjected to the requirements of Appendix J of 10 CFR 50, and the TS should be modified accordingly.1 2.6 Containment Pressure Setpoint (II.E.4.2.5)

The containment pressure setpoint that initiates containment isolation must be reduced to the minimum compatible with normal operating conditions. Most plants provided justification for not changing their setpoint and the NRC has approved their justification by separate 3

correspondence. The remaining plants must submit a change to the TS with the lower containment pressure setpoint and provide justification if this setpoint is more than 1 psi above maximum expected containment pressure during normal operatjon.

2.7 Containment Purge Valves (II.E.4.2.6)

Model TS were sent separately to each plant as part of the overall containment purge review. These TS include the requirement that the containment purge valves be locked closed except for safety related activities, verified closed at least every 31 days, and be subjected to leakage rate limits.1 2.8 Radiation Signal on Purge Valves (II.E.4.2.7)

The containment purge valves must close promptly to reduce the amount of radiation released outside containment following a release of radioactive materials to containment. TS should include the requirement that at least one radiation monitor that automatically closes the purge valves upon sensing high radiation in the containment atmosphere be operable at all times except cold shutdowns and refueling outages. If not operable, either the plant should begin proceeding to cold shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or the purge valves should be closed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Model TS were provided in Standard Technical Specifications format for those plants that are using safety-grade components to satisfy the requirement.I 2.9 Upgrade Babcock and Wilcox (B&W) AFWS (II.K.2.8)

Additional long-term AFWS modifications were to be performed in conjunction with Generic Letter 82-16 Items 3 and 4 (2.3 and 2.4 above).

The TS implemented for Items 3 and 4 will also address the upgrade of the B&W AFWS; therefore no separate TS changes would be required for this item for the B&W plants.

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2.10 B&W Safety-Grade Anticipatory Reactor Trip (II.K.2.10)

Safety-grade turbine trip equipment initiating a reactor trip was to be implemented by the B&W designed plants as part of the TMI lessons learned. The licensee is to implement in the TS the trip setpoint, number of channels, trip conditions, minimal channels required for operation, applicable operation modes, actions -to be taken, surveillance required and any other requirements for safety-grade equipment.

2.11 B&W Thermal-Mechanical Report (II.K.2.13)

Licensees of B&W operating reactors were required to submit by January 1, 1981, an analysis of the thermal-mechanical conditions in the reactor vessel during recovery from small breaks with an extended loss of all feedwater. TS, if required, will be determined following NRC staff

1 review.

2.12 Reporting Safety and Relief Valve Failures and Challenges (II.K.3.3)

NUREG-0660 stated that safety and relief valve failures be reported promptly and challenges be reported annually. The sections of the TS that discuss reporting requirements should be changed accordingly. The NRC has noted that an acceptable alternative would be to report challenges monthly.1 2.13 Anticipatory Trip on Turbine Trip (II.K.3.12)

Licensees with Westinghouse-designed operating plants have confirmed that their plants have an anticipatory reactor trip upon turbine trip.

Many of these plants already have this trip in the TS.

For those that do not, the anticipatory trip should be added to the TS.I 5

For HBR-2 the above Items 2.9, 2.10, and 2.11 are not being evaluated. Being a Westinghouse design, Items 2.9, and 2.10 are not applicable for HBR-2.

For Item 2.11, the Thermal-Mechanical report is being evaluated by the NRC staff as a separate active Three Mile Island (TMI) action item.

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3.

EVALUATION The evaluations of Generic Letter 82-16 Items are as follows:

3.1 STA Training (I.A.1.1.3)

The licensee stated that amendment 59 to the HBR-2 license, which was issued on August 24, 1981, included the qualification and training requirements for STA's. A review of the licensee TS5 shows the STA position is a part of the shift complement in Figure 6.2-2. Section 6.3.3 of the TS states that the STA shall have a bachelor's degree or equivalent in a scientific or engineering discipline, with specific training in plant design, and in the response and analysis of the plant for transients and accidents. The exact training program is not in the TS; however, the retraining and replacement program is covered in Section 6.4.1 of the TS.

In a letter from the NRC to CP&L, dated January 15, 1982,6 the NRC provided a post-implementation review of the CP&L STA training program.

The NRC concluded that the CP&L STA training program is acceptable in meeting the intent of the guidelines set forth. Until further guidance is issued by the Commission, no further licensing action is required for this item.

3.2 Shift Manning--Overtime Limits (I.A.1.3.1)

The licensee's response to this item states that "the CP&L policy regarding nuclear power plant staff working hours, which was submitted to the NRC, February 26, 1981, and approved by the NRC in a letter, dated November 5, 1981, is consistent with the intent of the commission policy as clarified in Generic Letter 82-12. The difference is that individuals are not permitted by CP&L policy to work more than 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months straight, rather than 16 and they are allowed to work 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months in a seven day period, rather than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ."

The policy is included in the CP&L administrative procedures rather than the TS.

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Since the CP&L policy is a deviation to the NRR criteria of Generic Letter 82-12, we recommend this item be evaluated as a separate issue.

Technical Assignment Control System (TAC) number 44124 has been assigned for active action of.this item.

3.3 Short Term Auxiliary Feedwater System (AFWS) Evaluation (II.E.1.1)

The licensee's response to this item states that an application to amend the existing operability requirements for the AFW system was made October 1, 1982. By a letter to CP&L, dated January 6, 1983,7 the NRC issued amendment number 74 for HBR-2 and a safety evaluation of the AFW system which found that the licensee has met the requirements of NUREG-0737, Item II.E.1.1-. Our review of the-HBR-2 TS, Sections 3.4 and 4.8, indicate that the limiting conditions for operation and the surveillance requirements are similar to the other related safety systems specified in the HBR-2 TS, thereby complying with the requirements of Generic Letter 82-16. No further licensing action is necessary for this item.

3.4 Safety Grade AFW Initiation and Flow Indication (II.E.1.2)

By a letter to CP&L, dated January 6, 1983,7 which discussed both the II.E.1.1 and II.E.1.2 issues, the NRC stated that open items of the original Safety Evaluation concerning safety grade flow indication and automatic initiation are under review and will be provided as a separate evaluation under TMI Item II.E.1.2 of NUREG-0737. Further licensing action may be required for this item following the NRC review. TAC number 44689 has been assigned for AFW Safety Grade automatic initiation and TAC number 44728 for the AFW Safety Grade Flow indication.

3.5 Dedicated Hydrogen Penetrations (II.E.4.1)

The licensee's response states, "The dedicated hydrogen penetration for HBR-2 was provided via an existing containment penetration, that no modifications or additional valves are required for the containment penetration; therefore, no valves need to be added to the TS for Appendix J testing."

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The licensee TS do not contain information which relates specifically to the hydrogen penetrations or Appendix J testing. FSAR Chapter-6, Table 6.2.4.1, lists the containment piping penetrations and valving, and identifies the containment vent system valves which are used for post accident gas control of the containment atmosphere. TS paragraph 4.4.2 and FSAR Chapter 6 paragraph 6.2.6.3 indicate that the isolation valves are subject to testing requirements of "1O CRF 50, Appendix J. We conclude that the Generic Letter 82-16 requirements for this item have been satisfied and no further licensing is required.

3.6 Containment Pressure Setpoint (II.E.4.2.5)

The licensee's response states that justification for not changing the containment pressure setpoint for HBR-2 had been found acceptable by the NRC, as indicated in a letter to CP&L dated December 21, 1981.

The licensee further stated that CP&L has discovered an error in the calculation of the containment net free volume used in the FSAR.

Subsequently, CP&L has determined that a TS change is required and submitted a TS change request on January 1, 1983, to correct the containment pressure setpoint. This item is being evaluated by the NRC as an action item under TAC number 51796. Further licensing action will be required for this item.

3.7 Containment Purge Valves (II.E.4.2.6)

The licensee has stated in his response "The containment purge valves for HBR-2 are locked closed with their breakers racked out and cleared.

Equipment that is cleared is verified to have the correct status every month by the Shift Foreman. The containment purge valves are continuously monitored for leakage via the Penetration Pressurization System (PPS).

Indication for excessive valve leakage would be indicated in the control room via the PPS alarm. Sufficient administrative controls are in effect and enforceable, therefore, no additional guidance is necessary in the TSs."

Review of the TS for HBR-2 indicates that there is no requirement in the TS for having the containment purge valves locked closed except for.

9

safety related activities and there is no surveillance requirement that the valves be verified closed at least every 31 days and subjected to leakage rate limits.

The requirements for NUREG-0737 Item II.E.4.2.6 are not met in the HBR-2 TS. Further licensing action will be required for this item.

3.8 Radiation Signal on Purge Valves (II.E.4.2.7)

The licensee has stated in his response, "Although the containment purge valves are locked closed with their breakers racked out, the respective radiation monitors and associated trips are functionally checked every two weeks. These radiation monitors are not in the TSs at the present time, but are incorporated in the Radiological Effluent Technical Specification (RETS) which will be submitted to the NRC in the near future." The licensee submitted the RETS to the NRC for review February 7, 1983 and it is presently being reviewed.

The requirements for NUREG-0737 Item II.E.4.2.7 have not been met in the HBR-2 TS. Further licensing action may be required for this item, following the NRC review.

3.9 Upgrade Babcock and Wilcox (B&W) AFWS (II.K.2.8)

H. B. Robinson Unit 2 is a Westinghouse design and, therefore, the requirements of this item are not applicable. No licensing action is required.

3.10 B&W Safety-Grade Anticipatory Reactor Trip (II.K.2.10)

H. B. Robinson Unit 2 is a Westinghouse design and, therefore, the requirements of this item are not applicable.. The anticipatory trip is evaluated under NUREG-0737 Item II.K.3.3 for the Westinghouse design. No further licensing action is required.

10

3.11 B&W Thermal-Mechanical Report (II.K.2.13)

H. B. Robinson Unit 2 is a Westinghouse design and, therefore, the requirements of this item are not applicable. It should be noted that there is a TMI action item, Thermal-Mechanical Report, for HBR-2 under TAC 46901. No further licensing action is required by Generic Letter 82-16 for this item.

3.12 Reporting Safety and Relief Valve Failures and Challenges (II.K.3.3)

The licensee has stated in his response that the HBR-2 Emergency Response Plan requires that the failure of a pressurizer safety or relief valve to close be declared an unusual event, and unusual events are required by 10 CRF 50.72 to be reported within one hour via telcon to the NRR. The licensee also stated that their administrative instructions require challenges to pressurizer PORV's and safety valves be included in the HBR-2 Annual Operating Report. It is CP&L's position that incorporating the reporting requirements into the TS would be redundant and unnecessary.

Review of TS Section 6.9.2 on reportable occurrences, concludes that the section does not specifically require that safety and relief valve failures and challenges be reported. Reporting of these failures is dependent entirely on other procedures. Unless a recent TS change has been submitted by CP&L or an acceptance of their policy has been issued by the NRC, further licensing action is required.

3.13 Anticipatory Trip on Turbine Trip (II.K.3.12)

The licensee has stated in his response that, "In a letter dated June 27, 1980, CP&L informed the NRC that HBR-2 has an at-power reactor trip for a turbine trip. The surveillance requirements for the turbine trip are included the TSs. The allowable setpoints for the turbine trip signal for reactor trip are not incorporated in the HBR-2 TS, because these setpoints are not used in the transient and accident analysis."

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Review-of the licensee TS shows that the limiting operating conditions for turbine trip and surveillance testing are contained in Tables 3.5-2 and 4.1-1, respectively. However, since the turbine trip signal is not included in the protection instrumentation settings for reactor trip, the licensee has not met the requirements of Generic Letter 82-16, and, unless a recent TS change has been submitted by CP&L or an acceptance of their policy has been issued by the NRC, further licensing action is required for this item.

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4. CONCLUSIONS Based on our review, we fi.nd the licensee conforms to those issues addressed in Generic Letter 82-16 on TS, except for those identified as follows:

1. Section 3.1 STA Training--Until further guidance is provided by the Commission, no further licensing action can be taken to determine whether the exact training program for the STA is required to be in the TS.

2. Section 3.2 Shift Manning-Overtime Limits--The CP&L policy is contained in administrative procedures rather than in the TS and, as stated in their response, the CP&L policy deviates from the NRR criteria.

3. Section 3.4 Safety Grade AFW System Initiation and Flow Indication--The NRC is presently reviewing open items of the original Safety Evaluation pertaining to safety grade flow indication and initiation as a separate evaluation.

4. Section 3.6 Containment Pressure Setpoint--The licensee has submitted a TS change request to correct the containment pressure setpoint. This request is presently under review by the NRC.

5. Section 3.7 Containment Purge Valves--The TS for H. B. Robinson Unit 2 do not comply with Generic Letter 82-16 for this item.

6. Section 3.8 Radiation Signal on Purge Valves--The TS for H. B. Robinson Unit 2 do not comply with Generic Letter 82-16 for

. this item. A licensee submittal for this is presently under NRC review and licensing action may be required, following this review.

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7.

Slection 3.12 Reporting Safety Valve and Relief Valve Failures and Challenges--The TS for H. B. Robinson Unit 2 do not comply with Generic Letter 82-16 for this item.

8. Section 3.13 Anticipatory Trip on Turbine Trip--The TS for H. B.. Robinson Unit 2 do not comply with Generic Letter 82-16 for this item.

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5. REFERENCES

1. D. G. Eisenhut, NRC letter to All Pressurized Power Reactor ticensees, "NUREG-0737 Technical Specifications (Generic Letter 82-16),"

September 20, 1982.

2. NUREG-0737 Clarification of TMI Action Plan Requirements published by the Division of Licensing, Office of Nuclear Reactor Regulation, U.S.

Nuclear Regulatory Commission, November 1980.

3. S. R. Zimmermann, Carolina Power & Light Co. letter to Darrell G. Eisenhut, Office of Nuclear Reactor Regulation, "H. B. Robinson Steam Electric Plant, Unit No. 2 Docket No. 50-261, License No. DPR-23, Response to Generic Letter No. 82-16, NUREG-0737 Technical Specifications," December 23, 1982.

4. D. G. Eisenhut, NRC letter to All Licensees of Operating Plants, Applicants for an Operating License, and Holders of Construction Permits, "Nuclear-Power Plant Staff Working Hours (Generic Letter No. 82-12)," JuneA15, 1982.

5. -H.

B. Robinson Steam Electric Plant, Unit No. 2 Technical Specifications, Appendix "A" to License No. DPR-23, Amendment No. 74, January 6, 1983.

6. Steven A. Varga, NRC letter to J. A. Jones, CP&L Company, "NUREG-0737 Item I.A.1.1 Shift Technical Advisor (STA)," January 15, 1982.

7. Glode Requa, NRC letter to E. E. Utley, CP&L Company, "Amendment No. 74 to Facility Operating License No. DPR-23 for H. B. Robinson Steam Electric Plant, Unit No. 2," January 6, 1983.

8. Steven A. Varga, NRC letter to J. A. Jones, CP&L Company, "Containment Setpoint Pressure TMI Task Action Plan Item II.E.4.2 Position 5 of NUREG-0737," December 21, 1981.

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NRC FORM 335

1. REPORT NUMBER IAssipnedby DDCJ U.S. NUCLEAR REGULATORY COMMISSION BIBLIOGRAPHIC DATA SHEET EGG-EA-6437

4. TITLE AND) SUBTITILE

2. (Leave bluki Conformance to NRR Generic Letter 82-16

3. RECIPIENT'S ACCESSION NO.

H. B. Robinson Steam Electric Plant Unit 2

7. AUTHOR(SI

5. DATE REPORT COMPLETED MONTH YE AR D. M. Beahm December 1983

9. PERFORMING ORGANIZATION NAME AND MAILING ADDRESS linclude Ztp Code)

DATE REPORT ISSUED MONTH YEAR December 1983 EG&G Idaho, Inc.

6 (Leave Nank)

Idaho Falls, ID 83415

8. (Leave bfank)

12. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS (include Zip Code)

10. PROJECT/TASK/WORK UNIT NO.

Division of Project and Resident Programs Region I

11.

FIN NO.

U.S. Nuclear Reguilatory Commission 101 Marietta Street, Suite 2900 A6600 Atlanta, Georgia 30303

13. TYPE OF REPORT PERIOD COVERED (Inclusve dars)

15. SUPPLEMENTARY NOTES

14. (Leave o.'k)

16. ABSTRACT 1200 words or less)

This EG&G Idaho, Inc. report evaluates whether the designated Operating Reactor Plant has conformed to the requirements of the NRR Generic Letter No. 82-16, "NUREG-0737 Technical Specifications."

17. KEY WORDS AND DOCUMENT ANALYSIS 17a. DESCRIPTORS 17b. IDENTIFIERS.OPEN-ENDED TERMS

18. AVAILABILITY STATEMENT

19. SECURITY CLASS (Ths report) 21 NO OF PAGES Unclassified Unlimited

20. CURITY CLASS (Thes page) 22 PRICE Unclassified NRC FORM 335 1tosi3

ML14183A025

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