IR 05000277/1982023
| ML20028F398 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 01/14/1983 |
| From: | Eapen P, Haverkamp D, Dante Johnson, Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20028F389 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-1.A.2.1, TASK-2.B.2, TASK-2.B.4, TASK-TM 278-82-22, 50-277-82-23, 50-278-82-22, NUDOCS 8302010325 | |
| Download: ML20028F398 (13) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-277/82-23 50-278/82-22 Docket No.
50-277 & 50-278 License No. DPR-44 & DPR-56 Priority
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Category C&C Licensee:
Philadelphia Electric Company 2301 Market Street Philadelphia, Pennsylvania 19101 Facility Name:
Peach Bottom Atomic Power Station, Units 2 and 3 Inspection At:
Delta, Pennsylvania and' Philadelphia, Pennsylvania Inspection Conducted: November 16-18, and December 8, 1982 Inspectors:
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f.3 D. JohnsorL/ Project EnginFer date' signed h
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P. Eapen, Reactor Engineer date signed bl2LE-ik In D. Haverkamp, Reacto censing Engineer date signed Accompanying NRC Personnel:
E. Tourigny, Lead Project Manager, Operating Reactors Branch 3, Division of Licc: sing, Office of Nuclear Reactor Regulation Approved by:
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E. McCabe, Chief, Reactor Projects Section date signed i
No. 2B, Division of Project and Resident Programs Summary:
November 16-18, and December 8, 1982 (Combined Report 50-277/82-23; 50-278/82-22)
Areas Inspected:
Special region-based inspection (54 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br />) of licensee action on a previous inspection finding and on implementation of NUREG-0737 Items 1.A.2.1.4, II.B.4.1 and II.B.2.
No violations were identified.
8302010325 830117 PDR ADOCR 05000277 Q
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DETAILS 1.
Persons Contacted Philadelphia Electric Company
- W. Alden, Engineer-In-Charge, Nuclear Section
- W. Birely, Senior Engineer-Licensing
- R. Bulmer, Superintendent, Nuclear Training Section
- M. Cooney, Superintendent, Generation Division-Nuclear A. Hillsmeier, Senior Health Physicist
- E. Purdy, Senior Engineer, Mechanical Engineering
- L. Pyril, Supervising Engineer, Mechanical Engineering
- W. Ullrich, Station Superintendent
- A. Wasong, Test Engineer H. Watson, Chemistry Supervisor
- J. Winzenried, Technical Engineer
- present at exit interview on December 8, 1982 2.
Licensee Action on a Previous Inspection Finding (0 pen) Unresolved Item 277/81-20-04; 278/81-22-05:
Loss of instrument air control manipulation required by requalification training program description, but not being performed during simulator training. Details are discussed in paragraph 4.
3.
Plant Shielding Design Review a.
Background and Scope A> discussed in Item II.B.2 of NUREG-0737, " Clarification of TMI Action Plan Requirements," each power reactor licensee was required to perform a radiation and shielding design review of spaces around systems that may, as a result of an accident, contain highly radio-active materials. The design review was intended to identify the location of vital areas * and equipment in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by the radiation fields during post-accident operatiors of these systems.
Additionally, each licensee was required to provide for adequate access to vital areas and protection of safety equipment by design changes, increased permanent or temporary shielding, or post-accident procedural controls.
The design review was to determine which types of corrective actions were needed for vital areas throughout the facility.
- Any area which will or may require occupancy to permit an operator to aid in the mitigation of or recovery from an accident is designated as a vital are.
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These requirements were discussed in NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Requirements"; were issued by NRC letters dated September 13 and October 30, 1982 to all operating nuclear power plants; and were incorporated into NUREG-0660,
"TMI-2 Action Plan." Significant changes in requirements or guidance were described in an NRC letter to all licensees of operating plants dated September 5, 1980, and were subsequently described in Item II.B.2 of NUREG-0737.
Lastly, an NRC letter to all licensees of operating power reactors dated March 17, 1982 (Generic Letter No. 82-05) requested reconfirmation of the schedule for completing Item II.B.2 of NUREG-0737.
The October 30, 1979 NRC letter indicated that licensee's plant shielding design reviews were among those items for which post-implementation NRC review is acceptable. Although prior NRC approval was not required, licensees were to document their methods of imple-mentation by the required completion date, e.g., design review by January 1, 1980 and plant modifications January 1, 1981.
With respect to documentation specified by NUREG-0737 for vital area access, operating license applicants were to provide to the NRC a sum-mary of the shielding design review, a description of the results of this review, and a description of the modifications made or to be made to implement the result of the review. The submittals were to include:
(1) Specification of source terms used in the evaluation, including time after shutdown that was assumed for source terms in systems; (2) Specification of systems assumed in the analysis to contain high levels of radioactivity in a post-accident situation; (3) Specification of areas where access is considered necessary for vital system operation after an accident; and, (4) The projected doses to individuals for necessary occupancy times in vital areas and a dose rate map for potentially occupied areas.
NUREG-0737 did not state that licensees of operating reactors were to submit the above documentation to the NRC.
Rather, they were to have available for review the final design details of the implementation *
of the Item II.B.2 position and clarification.
(Information equiva-lent to that submitted by operating license applicants is expected to be available for review as documentation of the design review that provided the bases for final design details.)
If deviations to that position and clarification were necessary, licensees were to provide detailed explanation and justification for the deviations by January 1, 1981.
- In addition to providing clarification of requirements, NUREG-0737 revised the completion date for modifications resulting from the plant shielding design review to January 1, 198 o
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The licensee's plant shielding design review and corrective actions were reviewed during this inspection. This review included licensee submittals to the NRC, a sampling verification of the shielding design review methodology and representative calculations, a review of selected emergency procedures to determine if the vital areas where personnel must go are safely accessible, and a review of corrective actions taken or planned by the licensee (including plant modifications),
b.
Licensee Submittals to the NRC In the case of Peach Bottom Atomic Power Station, Units 2 and 3, the shielding design review and corrective actions were discussed by the licensee in a letter to the NRC dated January 2, 1980. The licensee subsequently discussed the design review status and design details for the modifications in letters to the NRC dated January 31, 1980, October 15, 1980, January 8,1981, April 15,1982, and July 7,1982.
The information provided by these licensee sub!nittals, with respect to the plant shielding design review, is summarized below.
(1)
S. Daltroff letter to H. Denton dated January 2, 1980 (Subject:
Design Review Studies Required by Short Term Lessons Learned)
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The plant shielding design review was substantially com-pleted and identified modifications would be completed by January 1, 1981.
With respect to areas requiring continuous occupancy, the
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shielding design review included the Control Room and the Technical Support Center (TSC). No modifications were found necessary for the Control Room, however, this area was still under evaluation. One modification was needed for the TSC:
shielding would be added to the Unit 1 Control Room (the designated TSC) wall on the side facing the Unit 2 Reactor Building to bring down the dose from direct shine. This modification is discussed further in paragraphs 3.b.(2) and 3.e.
With respect to areas requiring infrequent access, the
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shielding design review included the Secondary Containment (which contains motor control centers, sample stations, instruments that measure reactor water level, and rooms that contain ECCS equipment), the Radwaste Building (which contains radwaste panels), and the Turbine Building (which contains the radiochemical laboratory). Modifications were not considered necessary based on the licensee's evaluation of post-accident operational requirements for access to motor control centers and rooms containing ECCS equipment. Modifications were being evaluated regarding the capability of obtaining primary coolant and primary containment samples, as discussed in the licensee's response to NUREG-0578, Section 2.1.8.a.
Modifications for access
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to the radwaste panel and the radiochemical laboratory were also being evaluated and dose calculations and determination of corrective action were expected to be completed by January 31, 1980.
With respect to instruments that measure water level, the
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licensee's January 2, 1980 letter stated:
" General Electric Company is evaluating the effects of an accident on the reactor vessel level instrumentation as part of NUREG-0578, Item 2.1.3.b.
Included in this evaluation is the determination of whether access is required to the reactor vessel level instrument racks to backfill the instrument line for the reference leg of the instrumentation. This evaluation is expected to be complete at the end of the year.
If access is required, we will provide a means of backfilling which is operable from an accessible area. This modification, if necessary, will be completed prior to January 1, 1981."
This matter is duscussed further in paragraph 3.d., and the inspector's findings are described in paragraph 3.f.(1).
(2)
S. Daltroff letter to H. Denton dated January 31, 1980 (Subject:
Design Review Studies Required by Short Term Lessons Learned)
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An evaluation to determine the airborne dose in certain plant areas had been completed. Additional corrective action was proposed.
In addition, the results of the Reactor Building dose assessment presented in the January 2, 1980 letter had been re-evaluated based on additional information.
With respect to areas requiring continuous occupancy, the
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post accident dose to personnel in the Control Room was confirmed to be within acceptable limits and no modifica-tions were necessary. However, shielding was still needed for the TSC and would consist of eight inches of concrete or equivalent.
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With respect to areas requiring infrequent access,' revised calculations indicated that secondary containment would be inaccessible for several days, and two modifications must be made regarding (1) the capability to obtain post-accident primary coolant and primary containment samples, and (2) con-trols and instrumentation associated with the makeup water supply to the spent fuel pools to permit maintenance of pool water level from outside secondary containment. Both modifi-cations were to be completed by January 1,1981, unless precluded by equipment unavailability.
(These modifications are dis-cussed further in paragraphs 3.f.(2) and 3.f.(3).) The dose calculations regarding access to the radwaste panel and radio-chemistry laboratory indicated no shielding modifications were required for personnel protection.
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(3)
S. Daltroff letter to D. Eisenhut dated October 15, 1980 (Subject: Implementation of NRC Action Plan Requirements)
Presented an assessment of the licensee's capabilities
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to implement near term NUREG-0660 requirements and a proposed schedule for implementation in response to the September 5, 1980 NRC letter.
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With respect to the plant shielding study, the licensee discussed the relocation of facilities and equipu nt, pro-posed for completion by January 1, 1931 (ss presented in the January 31, 1980 submittal) and specifically noted this 1,volves relocation of the spent fuel maieup controls to areas outside the Reactor Building. The licensee stated further that an NRC Region I meeting held in Arlington, Virginia on September 22, 1980 provided additional clarifi-cation of the source term design criteria for the plant shielding study. The licensee's reassessment of the shielding study, based on this new clarification, indicated that post-accident radiation conditions will not impact on reactor building accessibility. Therefore, the licensee proposed that implementation of the modifications described above be deferred until such time that their need is clearly established.
(This modification is discussed further in paragraph 3.f.(3).)
(4)
S. Daltroff letter to D. Eisenhut datei January 8, 1981 (Subject:
Information Reouested by NUREG-0737)
Responded to sever 11 of the TMI related NUREG-0737 requests
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that licensees submit specific information regarding plant systems and equipment or the results of engineering studies evaluating new design standards.
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Although NUREG-0737 Item II.B.2 required no responses from licensees of operating reactors (unless deviations to the position or clarification were necessary), this submittal stated, in part:
" Based upon the clarified source term design criteria and the expanded vital area criteria of NUREG-0737, the results presanted in our submittal of January 31, 1980, S. L. Daltroff to H. R. Denton, indicate that the post-accident radiation conditions will not impact on accessibility to vital areas defined for PBAPS (Peach Bottom Atomic Power Station)."
The inspector noted, during discussions with licensee management, that the above information did not appear to be supported by an adequate evaluation of all potentially vital areas identified in NUREG-0737. This matter is discussed further in paragraphs 2.c and 3.f.(2).
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(5)
S. Daltroff letter to D. Eisenhut dated April 15, 1982 (Subject:
Response to Generic Letter 82-05)
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With respect to the plant shielding design review, the submittal stated that the study was completed. The licensee concluded that current design provides access to vital areas under accident conditions (submittals dated January 2, 1980, January 31, 1980, October 15, 1980, January 8,1981), therefore, modifications are not deemed to be necessary.
The inspector noted that this information is not totally correct, in that certain aspects of the shielding design review are incomplete, including evaluation of the need to complete modifications previously considered necessary.
This matter is discussed further in paragraphs 3.c, 3.f.(1),
and 3.f.(3).
Also, shielding was provided in the TSC in accordance with the licensee's earlier commitments, as discussed in para-graph 3.e.
(6)
S. Daltroff letter to D. Eisenhut dated July 7, 1982 (Subject:
NUREG-0737, Item II.B.2, Plant Shielding Evaluation)
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Provided additional information regarding the previously submitted shielding studies, including identification of methodologies and assumptions used in the Peach Bottom shielding study that may be inconsistent with some of the NRC guidance.
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Information provided was related to studies performed for l
the Control Room and for emergency response facilities (Technical Support Center and Emergency Operations Facility).
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Based on the results of the revised shielding study, no additional modifications were identified as being needed with respect to plant shielding.
The inspector noted that with respect to the shielding design review guidance described in NUREG-0737, Item II.B.2, the licensee's methodologies and assumptions'are considered acceptable. However, the information in this submittal is also related to NUREG-0737, Item III.D.3.4, Control Room Habitability Requirements, and NUREG-0696, Functional Criteria for Emergency Response Facilities. The submittal was not reviewed by the inspectors with respect to those items, which are being reviewed separately by the NRC staff.
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c.
Shielding Design Review Veri *ication The licensee's shielding design review methods, including source terms, assumptions and bases, calculation of dose rates, calculation of total integrated doses (TID's)-for 180 days, and results, were described in Document No. 010878,-" Evaluation of Control Room Habit-ability and Plant Shielding Review Following DBA-LOCA for the Peach Bottom Atomic Power Station," prepared by Bechtel Power Corporation.
The licensee described various aspects of this report in submittals to the NRC, as discussed in paragraph 3.b.
The inspector discussed the details of the ' shielding design review with representatives of the licensee and his contractor. These individuals provided additional information regarding the assumptions
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and methodology used in dose rate calculations and the results of
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such calculations. The assumptions were consistent with the guide-lines of NUREG-0737 Item II.B.2.
The methodology and mathematical model employed state-of the-art techniques for shielding design. The inspector compared the dose rates obtained from the licensee's calcu--
lations with those developed by NRC consultants for a similar plant configuration, and the dose rates were consistent. The licensee's shielding design review methodology and dose rate calculations were acceptable, and the inspector had no further questions in these areas.
The licensee's evaluation of plant shielding, as described in Document No. 010878, provided extensive information with respect to dose rate calculations in many plant areas and general information regarding systems assumed to contain high levels of radioactivity in a post-accident situation. However, neither this document nor other licensee documentation provided specification of areas where access is considered necessary for vital system operation after an accident, or an evaluation of all potentially vital areas discussed in NUREG-0737, Item II.B.2.
In addition, the licensee did not have documenta-tion that described the projected doses to individuals for necessary occupancy times in vital areas. During discussions with site and corporate office engineering staff personnel and with licensee manage-ment, the inspector determined that vital areas were identified and-dose levels were calculated during 1979'in response to NUREG-0578,-
Item 2.1.6.b.
This information may have been informally documented.
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This is supported by the fact that some vital areas were identified and some personnel dose estimates were discussed in various licensee submittals. However, this information apparently was not substan-tiated by appropriate licensee documentation as discussed below.
With respect to vital area identification, the clarification of NUREG-0737, Item II.B.2 states:
"The Control Room, Technical Support Center (TSC), sampling station and sample analysis area must be included among those areas where access is considered vital after an accident....
The evaluation to determine the necessary vital areas should also include, but not be
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limited to, consideration of the post-LOCA hydrogen control system,.
containment isolation reset control area, manual ECCS alignment area (if any), motor control centers, instrument panels, emergency power.
j supplies, security center, and radwaste panels...."
As stated previously, some of the above areas and other. areas were conveyed as vital areas, based on post-accident operational require-ments, in various licensee submittals. However, several of the areas identified for consideration in NUREG-0737 Item II.B.2 have not been evaluated formally by the licensee. Therefore,'the licensee's shielding design review is incomplete regarding the identification of-vital areas and determination of appropriate types of corrective actions needed to provide for adequate access to vital areas. This item is considered unresolved pending completion of licensee actions.
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(277/82-23-01/ 278/82-22-01).
d.
Vital Area Accessibility - Procedure Review The inspector reviewed two emergency procedures that would be imple-l mented by the licensee in the event of various severities of loss of coolant accidents. The review included (1) a plant walkdown of-portions of each procedure to determine the ability to perform the
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local operation, and (2) an assessment of potential exposures to plant personnel based on the results of,the licensee'.s shielding design review. The procedures reviewed included Emergency Procedure E-14 "Large Break-Loss of Coolant Accident - Offsite Power Available,"
Revision 15 dated May 17, 1982, and Emergency Procedure E-15 " Loss of
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Coolant Accident Concurrent with Loss of Offsite Power - Loss of All Seismic Class II Equipment - Failure of One Diesel Generator to-
Start," Revision 14 dated May 19, 1982.
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Followup Action step 14 of Emergency Procedure E-14 states:
" Notify (I&C) Lab to backfill.(reactor) level instrumentation lines.
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This will provide reliable reactor vessel level instrumentation."
The inspector noted that this action would be performed'at.the 165'
elevation of the Reactor Building, which may be inaccessible due'to
post-accident high radiation conditions. However, procedural con-
trols have not been established to provide the method-(pre planned-
access route, instructions for valve operations, etc.).for.backfilling
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the instrument lines.
The licensee's submittal to the NRC dated
January 2, 1980, stated that General Electric Company was evaluating the effects of an accident on reactor vessel instrumentation, includ-ing the determination of whether access is required to.the reactor vessel level instrument racks to backfill the instrument reference
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legs.
If access was required, a means of backfilling was to be provided by January 1, 1981. The licensee evaluation of this matter
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and determination of corrective actions has not been completed. This matter is discussed further in paragraph 3.f.(1).
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e.
Modifications and Corrective Actions Based on the results of the plant shielding design review, the licensee deterreined that the calculated doses would preclude con-tinuous post-ar.cident access needed for the Technical Support Center (TSC).
Therefore, the licensee committed to installing shielding (eight inches of concrete or equivalent) on the side of the TSC that faced the Unit 2 Reactor building, as discussed in paragraphs'3.b.(1)
and 3.b.(2).
The inspector verified that the shield wall was installed per the licensee's commitments and had no further questions regarding this matter.
f.
Findings
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(1) As described in paragraph 3.b.(1), the licensee committed (January 2,1980 submittal) to evaluating the need for access to backfill reactor vessel instrument lines and, if necessary, to provide a means for backfilling from an accessible area. As noted in paragraph 3.b.(5), the licensee subsequently concluded (April 15, 1982 submittal) that the current design provides ade-quate acess to vital areas, however, the evaluation of backfilling instrument lines was not specifically discussed. As discussed in paragraph 3.d., Emergency Procedure E-14 specifies backfilling the instrument lines as a followup action for a loss of coolant acci-dent, however, no provisions have been included for performing this operation from an accessible area.
Licensee evaluation of backfilling the instrument lines and determination of appropriate corrective actions (design change, increased permanent or temporary shielding, or post-accident procedural controls) is considered part of the unresolved item discussed in paragraph 3.c. (277/82-23-01; 278/82-22-01).
(2) As described in paragraph 3.b.(2), the licensee committed (January 31, 1980 submittal) to completing a modification regarding the capability to obtain post-accident primary coolant and primary containment samples. This modification is related to NRC requirements described in NUREG-0737, Item II.B.3,
" Post-Accident Sampling Capability," which was not included within the scope of this inspection. The modification is subject to review during a subsequent NRC inspection.
(3) As described in paragraph 3.b.(2), the licensee committed-(January 31, 1980 submittal) to completing a modification regard-ing the controls and instrumentation associated with the make-up water supply to the spent fuel pools to permit maintenance of water level from outside secondary containment. As noted in paragraph 3.b.(3), the licensee proposed (October 15, 1980) that implementation of this modification be deferred until such time as the need is clearly established. The basis for deferral, as stated by the licensee, was that reassessment of the shielding study, based on additional clarification of the source term
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design criteria provided during a September 22, 1980 meeting with the NRC, indicated that post-accident radiation conditions will not impact on reactor building accessibility.
The "addi-tional clarification" which led to the licensee's conclusion was not described further in the licensee's submittal. The licensee's shielding design review discussed in paragraph 3.c. does not support this conclusion, in that data for several areas of the reactor building indicate very high post-accident dose rates due to equipment / piping shine. Based on the inspector's review of the shielding design review data, the licensee's general state-ments (January 8, 1981 submittal) that " post-accident radiation conditions will not impact on accessibility to vital areas defined for PBAPS," as discussed in paragraph 3.b.(4), and (April 15, 1982 submittal) that the " current design provides access to vital areas under accident conditions," as described in paragraph 3.b.(5), also appear to be unsupported. The licensee's specific evaluation of completing a modification to permit post-accident maintenance of spent fuel pool water level from outside secondary containment and the general evaluation that current design provides access to vital areas is. considered part of the unresolved item discussed in paragraph 3.c. (277/
82-23-01; 278/82-22-01).
4.
Training and Requalification The inspector reviewed the licensee's conformance with requirements and I
guidance described in NUREG-0737, Items I.A.2.1, "Immediate Upgrading of Reactor Operator and Senior Reactor Operator Training and Qualifications,"
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and II.B.4, " Training for Mitigating Core Damage." The review included an assessment of an NRC contractor's (Science Applications, Inc.) Technical
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Evaluation Report (TER), a review of applicable licensee submittals to the
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concluded that the licensee's training and requalification programs satisfied the NRC guidelines except for two areas.
In addition, one
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previous inspection finding (an unresolved item) remained open. The inspector discussed these items with onsite and corporate office licensee representatives, as described below.
The TER concluded that the licensee's initial and requalification training in mitigation of core damage and related subjects did not satisfy the NRC staff's acceptance criterion of 80 cortact hours. The "80-hour criterion" had been provided to the NRC's contractor by the staff as a review criterion for determination of training program acceptability based on the amount of training provided.
If a licensee's initial program included at least 80 contact hours of instruction, then the program would be considered adequate.
i However, if less than 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> were provided by a licensee, then more detailed information would be needed by the NRC staff to determine adequacy.
In the case of Peach Bottom Units 2 and 3, the training program for mitiga-tion of core damage and related subjects was reviewed during NRC Inspections 50-277/81-20 and 50-278/81-22 and found acceptable. The 80-hour review l
criterion was also originally applied by the NRC staff to acceptance reviews
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of requalification programs for licensed operators. However, the-staff ~
rescinded this criterion, as it was considered. inappropriate to' prescribe a generic duration for any portion of the licensed operator requalification program. Therefore, the inspector had no further questions in this area and noted that the contractor's TER conclusion would be clarified and resolved in the staff's safety evaluation of this matter.
NUREG-0737, Item II.B.4 states that " shift technical advisors =and operating personnel from the plant manager through the operations chain to the licensed operators" shall receive training on the use of installed systems to control or mitigate accidents in which the core is severely damaged.
The TER correctly noted, based on licensee submittals, that the Station Superintendent had not received this training. This matter was discussed with the Station Superintendent and the licensee's training manager, who provided the inspector with internal documentation certifying that the Station Superintendent has received sufficient training to meet the intent'
of NUREG-0737 Item II.B.4.
Licensee management stated that information will be submitted to the NRC that will confirm completion of this training, so that the staff's review of this matter can be completed. The_ licensee's actions regarding this item will be reviewed during a subsequent.NRC inspection.
(277/82-23-02; 278/82-22-02).
The previous unresolved item (277/81-20-04; 278/81-22-05) included a discrepancy between control manipulations identified in the licensee's program description-for licensed operator requalification training and the " Simulator Requalification Reactivity and-Control Manipulations: Record Sheet," which is used to document the performance of control manipulations.
Specifically, " loss of instrument air" is listed as a control manipulation in the program description, but the manipulation cannot be simulated at the Limerick Training Center (used for simulator requalification of-Peach Bottom operators) and, therefore, the manipulation is not included on the record sheet. This finding was identified during an NRC inspection conducted on September 4 - October 6, 1981. NUREG-0737 Item I.A.2.1 included a specific-listing of control manipulations to be performed during operator requalification programs. One of the manipulations listed'
was loss of instrument air (if simulated plant specific). As noted above, this manipulation is not simulated for Peach Bottom Units 2 and 3, and thus the manipulation is not required. However, the licensee's submittal to the NRC staff dated April 30, 1981, identified. loss of instrument air as a control manipulation inc'Jded in the simulator portion of.the requalifi -
cation program. That information was incorrect.
During the exit' meeting on December 8, 1982, licensee management stated that this matter would be clarified in a submittal to the NRC staff by December 24, 1982. This item is unresolved pending review of the licensee's submittal during a subsequent.
NRC inspection.
(277/82-23-03; 278/82-22-03).
During onsite review of the above item concerning the loss of instrument air control manipulation, the inspector determined that the licensee has no procedure for loss of instrument air. The facility Technical Specifi-cations Section 6.8.1 states, in part: " Written procedures and administra-tive policies shall be established, implemented and maintained that meet
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the requirements of Sections 5.1 and 5.2 of ANSI N18.7-1972 and Appendix
"A" of the USAEC Regulatory Guide 1.33 (November 1972)..." Loss of Instrument Air is listed by Regulatory Guide 1.33 (November 1972) as one of the procedures for combating emergencies and other significant events.
Licensee management noted that the need for a loss of instrument air procedure was discussed with NRC inspectors during the initial licensing phase of Peach Bottom Units 2 and 3.
At that time, the licensee's position was that no specific emergency procedure was required, apparently based on the inherent redundancy of the installed instrument air systems and the low probability of loss of instrument air.
Furthermore, there were no NRC requirements for the licensee to have this procedure at the time of licensing, and no procedure was established. The inspector noted'that the requirements for such a procedure were promulgated several years later.
And, because the licensee is revising the emergency procedures to be consistent with revised NRC guidelines for function-based emergency response procedures, no loss of instrument air emergency procedure may be necessary to assure safe operation. Nonetheless, based on literal Techni-cal Specification requirements, either a procedure is necessary or the Technical Specifications need changing.
Licensee management stated that this matter would be evaluated in accordance with Technical Specifica-tion requirements. This item is unresolved.
(277/82-23-04; 278/82-22-04)
5.
Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, violations or devia-tions. An unresolved item is discussed in paragraph 3.c. and 3.f.
Addi-tional unresolved items are discussed in paragraph 4.
6.
Exit Interview The inspector met with licensee representatives (denoted in paragraph 1)
at the conclusion of the inspection on December 8, 1982, to discuss the inspection scope and findings.
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