ML18082A678
| ML18082A678 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 06/27/1980 |
| From: | Mittl R Public Service Enterprise Group |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML18082A679 | List: |
| References | |
| NUDOCS 8007070285 | |
| Download: ML18082A678 (14) | |
Text
Public Service Electric and Gas Company 80 Park Place Newark, N.J. 07101 Phone 201/430-7000 June 27, 1980 Director of Nu.clear Reactor Regulation
- u. S. Nuclear Regulatory Commission Washington, D.C.
20555 Attention:
Mr. A. Schwencer, Acting Chief Licensing Branch No. 3 Division of Licensing Gentlemen:
FULL POWER LICENSE REQUIREMENTS NO. 2 UNIT SALEM NUCLEAR GENERATING STATION DOCKET NO. 50-311 PSE&G hereby submits, in the enclosure to this letter, its response to the full power license requirements identified at a meeting with members of the NRC staff on June 19, 1980.
Should you have any questions in this regard, please do not hesitate to contact us.
Encl.
M P80 64 09 15 The Energy People
_s_o _0_1 _o _7*_0_2_e_~ ___ p f!!JJ?rs, R. L. Mittl General Manager -
Licensing and Environment Engineering and Construction 95-0942
FULL POWER REQUIREMENTS NO. 2 UNIT SALEM NUCLEAR GENERATING STATION TMI-2 RELATED ITEMS
- l. c. 7 NSSS VENDOR REVIEW OF PROCEDURES NSSS vendor review of power ascension test and emergency procedures to further verify their adequacy.
Response
Westinghouse is in the process of reviewing the Salem Erner-gency Operating Procedures and Power Ascension Test Proce-dures.
This activity will be completed prior to escalation above 5% power.
- l. c. 8 PILOT MONITORING OF SELECTED EMERGENCY PROCEDURES FOR NEAR-TERM OPERATING LICENSE APPLICANTS Correct emergency procedures, as necessary, based on the NRC audit of selected plant emergency operating procedures (e.g., small-break LOCA, loss of feedwater, restart of engi-neered safety features following a loss of ac power, steam line break, or steam generator tube rupture).
Response
Selected Emergency Operating Procedures requested by the NRC staff will be submitted for review by July 25, 1980.
I.G.l TRAINING DURING LOW-POWER TESTING Supplement operator training by completing the special low-power test program.
Tests may be observed by other shifts or repeated on other shifts to provide training to the oper-ators.
M P80 64 09/l Response _
Operator training will be supplemented during the conduct of the Special Low Power Test Program, as described in PSE&G letter, R. L. Mittl to o. D. Parr, dated March 31, 1980.
II.B.l REACTOR COOLANT SYSTEM VENTS Provide a description of the design of reactor coolant sys-tem and reactor vessel head high point vents that are re-motely operable from the control room and supporting analy-ses.
This requirement shall be met before issuance of a full-power license.
(See Enclosure 4 to letter of September 27 and November 9, 1979.)
Response
This information was provided in PSE&G letter, R. L. Mittl to O. D. Parr, dated January 4, 1980.
II.B.2 PLANT SHIELDING Provide (1) a radiation and shielding design review that identifies the location of vital areas and equipment in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by radiation during oper-ations following an accident resulting in a degraded core, and (2) a description of the types of corrective actions needed to assure adequate access to vital areas and protec-tion of safety equipment.
(See NUREG-0578, Section 2.l.6b, and letters of September 27 and November 9, 1979.)
Response
This information was provided in PSE&G letter, R. L. Mittl too. D. Parr, dated January 4, 1980.
Additional informa-tion will be provided by July 1, 1980.
M P80 64 09/2
r II.B.3 POSTACCIDENT SAMPLING Provide (1) a design and operational review of the capability to promptly obtain and perform radioisotopic and chemical analyses of reactor coolant and containment atmos-phere samples under degraded core accident conditions with-out excessive exposure, (2) a description of the types of corrective actions needed to provide this capability, and (3) procedures for obtaining and analyzing these samples with the existing equipment.
(See NUREG-0578, Section 2.l.8a and letter of September 27 and November 9, 1979).
Response
This information was provided in PSE&G letter, R. L. Mittl to o. D. Parr, dated January 4, 1980.
Additional informa-tion will be provided by July 1, 1980.
II.B.4 TRAINING FOR MITIGATING CORE DAMAGE Complete the training of all operating personnel in the use of installed systems to monitor and control accident in which the core may be severely damaged.
Response
Initial training will be provided as described in PSE&G let-ter, R. L. Mittl to o. D. Parr, dated April 1, 1980.
II.E.l.l AUXILIARY FEEDWATER SYSTEM RELIABILITY EVALUATION (1)
Provide a simplified auxiliary feedwater system relia-bility analysis that uses event-tree and fault-tree logic techniques to determine the potential for AFWS failure following a main feedwater transient, with par-ticular emphasis on potential failures resulting from human errors, common causes, single point vulnera-bility, and test and maintenance.,outage.
(2)
Provide an evaluation of the AFWS using the acceptance criteria of Standard Review Plan Section 10.4.9.
M P80 64 09/3
r
.* (3)
Describe the design basis accident and transients and corresponding acceptance criteria for the AFWS.
(4)
Based on the analyses performed modify the AFWS, as necessary.
Response
The AFW systems for both Salem l and 2 are identical in design, including equipment and automatic initation logic.
The Technical Specifications for both units are 'essentially identical, with the exception that the Salem 2 Technical Specifications contain a monthly surveillance requirement for verification of discharge pressure for each motor driven AFW pump and an annual surveillance requirement to verify that the Service Water spool piece is on site.
A copy of the Salem l and 2 AFW Technical Specifications (with the differences noted) is provided in Attachment 1.
The AFW System reliability analysis that was performed for the Salem l AFW System by the NRC staff and documented in NUREG-0611, "Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse -
Designed Operating Plants," January, 1980, is also applic-able to the Salem 2 AFW System.
It should be noted that the two princip,al dependencies ide n-tified in Section X.12.2.2 of your letter of September 21, 1979, namely, the single manual valve (AFl) in the AFW sue-M P80 64 09/4 tion line to the AFST and the throttle-back practice used at that time have been eliminated through implementation of recommendations GS-2 and GS-3.
The elimination of these dependencies enhances the overall reliability of the AFW System.
II~E.3.1 EMERGENCY POWER FOR PRESSURIZER HEATERS Install the capability to supply from emergency power buses a sufficient number of pressurizer heaters and associated controls to establish and maintain natural circulation in hot standby conditions.
(See NUREG-0578, Section 2.1.1, and letters of September 27 and November 9, 1979.)
Response
This modification has been completed as described in PSE&G letter, R. L. Mittl to O. D. Parr, dated January 4, 1980.
II.E.4.2 CONTAINMENT ISOLATION DEPENDABILITY Provide (1) containment isolation on diverse signals, such as containment pressure or ECCS actuation, (2) automatic isolation of nonessential systems (including the bases for specifying the nonessential systems), (3) no automatic re-opening of containment isolation valves when the isolation signal is reset.
(See NUREG-0578, Section 2.1.4, and let-ters of September 27 and November 9, 1979.)
Response
This information was provided in PSE&G letter, R. L. Mittl to O. D. Parr, dated January 4, 1980.
Additional information will be provided by July 1, 1980.
II.K.3 C.3.3 FINAL. RECOMMENDATIONS OF B&O TASK FORCE Assure that any failure of a PORV or safety valve to close will be reported to the NRC promptly.
All challenges to the PORVs or safety valves should be documented in the annual report.
M P80 64 09/5
Response
Any failure of a PORV or safety valve to close will be promptly reported to the NRC.
All challenges to the PORVs or safety valves will be documented in the annual report.
III.A.l.l UPGRADE EMERGENCY PREPAREDNESS Provide an emergency response plan in substantial compliance with NUREG-0654, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants" (which may be modified after May 13, 1980 based on public comments) except that only a description of and completion schedule for the means for providing prompt notification to the population (App.
3), the staffing for emergencies in addition to that already required (Table B.l), and an upgraded meteorological program (App. 2) need be provided.
NRC will give substantial weight to FEMA findings on offsite plans in judging the adequacy against NUREG-0654.
Perform an emergency response exercise to test the integrated capability and a major portion of the basic elements existing within emergency preparedness plans and organizations.
Response
Additional information will be provided by July 1, 1980.
III.D.l.l PRIMARY COOLANT SOURCES OUTSIDE CONTAINMENT Reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as-low-as-practical levels, measure actual leak rates arid establish a program to maintain leak-age at as-low-as-praGtical levels and monitor leak rates.
(See NUREG-0578, Section 2.1.ba, and letters of September 27 and November 9, 1979.)
Response
This information was provided in PSE&G letter, R. L. Mittl M P80 64 09/6 to o. D. Parr, dated January 4, 1980.
Additional informa tion will be provided by July 1, 1980.
III.D.3.4 CONTROL ROOM HABITABILITY Identify and evaluate potential hazards in the vicinity of the site as described in SRP Sections 2.2.1, 2.2.2, and 2.2.3, confirm that operators in the control room are ade-quately protected from these hazards and the release of radioactive gases as described in SRP Section 6.4, and, if necessary, provide the schedule for modifications to achieve compliance with SRP Section 6.4.
Response
This information will be provided by July 1, 1980.
M P80 64 09/7 FULL POWER GENERIC ITEMS (NON-TMI)
- 1.
Environmental Qualification Comply with the guidelines of NUREG-0588.
Response
PSE&G has provided environmental qualification data applica-ble to Salem 2 in response to previous NRC Questions Q7.30, Q7.31, Q7.32, Q7.33, and Q7.35 in FSAR Amendment 43, and in the correspondence referenced in Attachment 2.
PSE&G's response to Bulletin 79-01, "Environmental Qualifi cation of class lE Equipment," for Salem 1, dated March 6, 1979, stated that a review was being performed on the Salem 2 License application and that the information generated would constitute the review for Salem 1.
Upon receipt of the NRC's October 11, 1979 letter, an expanded review and the preparation of data in the desired format was initiated with the intent of responding to the Salem 2 request, and as a supplemental response to the Bulletin.
However, in the course of review and preparation of a response, Bulletin 79-0lB was issued in January, 1980, which required a comprehensive review of equipment qualification in a different format than that requested in the October letter.
As a result of this Bulletin, PSE&G established a review plan to accomplish a qualification evaluation in such M P80 64 09/8 a manner that varied requirements could be addressed in an orderly fashion.
The Bulletin was applicable to operating plants; since Salem 1 and 2 are essentially identical, the qualification data developed for the Salem 1 review would also apply to Salem 2.
A different review requirement (NUREG-0588) of a much larger scope was requested for Salem 2 in the February 19, 1980 NRC letter.
The effort that was in progress to meet Bulletin 79-0lB requirements could accommodate changes in scope and requirements, but would result in a somewhat longer review time.
Since Salem 1 and 2 are essentially the same, the review has continued on that basis, with emphasis on Salem 1, to prepare the submittal forms required by the Bulletin.
The response for Salem 2 will be provided upon completion of the review.
It is expected that the information requested in Bulletin 79-0lB will be submitted in September, 1980.
M P80 64 09/9 In addition to Bulletin 79-0lB and NUREG-0588, the review will address item 2.l.6b of NUREG-0578.
We believe that continuation of the present review effort will be of benefit to both PSE&G and the NRC.
Since Salem l and 2 are essen-tially identical, compilation of data at one time with subsequent sorting and submittal of information as necessary to meet the different submittal requirements will assure a timely response for all requirements and a consistent, comprehensive evaluation of the information.
The data that has been submitted for Salem 2, as noted in, does address a portion of the concerns and requirements of NUREG-0588.
The NUREG-0588 response for safety-related equipment subject to high energy line breaks will be completed on a schedule consistent with Bulletin 79-0lB.
Due to the expanded scope of NUREG-0588, the information requested for safety-related equipment in a controlled environment may take longer to develop.
A list of submittals in response to Bulletin 79-0lB is also provided in Attachment 2.
M P80 64 09/10
- 2.
PAD-3.3 Performance Code Complete evaluation regarding a restriction in the use of this code.
Response
NRC item.
See SER Supplement 4, Section 4.2.2.
- 3.
ATWS Review and approve Operating Procedures.
Response
Operating procedures were submitted in PSE&G letter, R. L.
Mittl to O. D. Parr, dated March 13, 1980.
Currently await-ing NRC staff guidance regarding acceptability.
See SER Supplement 4, Section 7.2.2.
- 4.
Response
Response will be provided by July 1, 1980, in accordance with Confirmatory Order dated April 4, 1980.
See SER Sup-plement 4, Section 7.9.
- 5.
Diesel Generator Reliability Compliance with RG 1.108 and NUREG/CR-0660.
Response
NRC item.
See SER Supplement 4, Section 8.3.4.
M P80 64 09/11
- 6.
Topical Reports WCAP-9226, 9230 and 9236 related to main steam and feedline break accidents.
Response
NRC item.
Prompt responses will be provided to any NRC staff questions directed to PSE&G during the NRC's review of these Topical Reports.
See SER Supplement 4, Section 15.1.1.
- 7.
Q List Complete review of "Q List" requirements.
Response
NRC item.
See SER Supplement 4, Section 17.l.
- 8.
Response
PSE&G's response to IE Bulletin 80-06 was submitted by PSE&G letter, F. W. Schneider to B. H. Grier, dated June 13, 1980.
- 9.
Evaluation of Manning Requirements
Response
NRC item.
See SER Supplement 4, Section 13.l.l.
M P80 64 09/12 FULL POWER PLANT SPECIFIC ITEMS (NON-TMI)
- 1.
Response
Additional information will be submitted by July 1, 1980.
M P80 64 09 13