ML20154Q917
| ML20154Q917 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 03/20/1986 |
| From: | Taylor J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | Kingsley O MISSISSIPPI POWER & LIGHT CO. |
| References | |
| EA-84-075, EA-84-75, NUDOCS 8603240173 | |
| Download: ML20154Q917 (17) | |
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UNITED STATES
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[,f.i MAR 2 01996 Docket No. 50-416 License No. NPF-29 EA 84-75 Mississippi Power and Light Company ATTN: Mr. O. D. Kingsley, Jr.
Vice President-Nuclear Operations P. O. Box 1640 Jackson, Mississippi 39215-1640 Gentlemen:
This refers to the letter dated April 19, 1985 from Mississippi Power and Light Company (MP&L) to the Director, Office of Inspection and Enforcement in response to the Notice of Violation and Proposed Imposition of Civil Penalties sent to you on March 21, 1985. The letter and Notice described violations that occurre'd as a result of several licensee submittals made to the NRC between December 15, 1980 and August 5, 1984.
I have carefully considered your response and, after consultation with the Commission, have determined that Violations A, B, C, and D should be withdrawn or recategorized in whole or in part and the Twenty-Five Thousand Dollar ($25,000) penalty proposed for each of Violations A, B, C, and D of the Notice should be withdrawn. The basis for this determination is discussed in the enclosure to this letter. We have also determined that in view of the mistakes made in the other citations, no civil penalty should be imposed for the remaining violation, Violation E.
This completes the agency's action in this matter.- No further response is required.
In accordance with Section 2.790 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations, a copy of this letter and the enclosure will be placed in the NRC's Public Document Room.
Sincerely, g/
s M. Tay1
, Director fice of In ection and Enforcement
Enclosure:
Evaluation And Conclusions Regarding MP&L Response To EA 84-75
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EVALUATION AND CONCLUSION REGARDING MP&L RESPONSE TO EA 84-75 In Mississippi Power and Light Company's (MP&L's or licensee's) April 19, 1985 response to the Notice of Violation and Proposed Imposition of Civil Penalties dated March 21, 1985, the licensee requested that the NRC staff reconsider the appropriateness of or further mitigate the proposed civil penalties. A statement of each violation, a summary of the licensee's response, and the NRC's evaluations and conclusions are presented below:
Violation A Restatement of Violation i
A.
On December 15, 1980, a markup of the Standard Technical Specifications for General Electric Boiling Water Reactors (NUREG-123), Revision 2, 3
August 1979, was submitted. The transmittal letter contained the statement that the markup reflected plant specific design features for Unit 1 of the l
Grand Gulf Nuclear Station.
Contrary to Section 186 of the Atomic Energy Act of 1954, as amended, this statement was a material false statement. The statement was false because, in the following respects, the technical specifications did not reflect existing plant specific design features.
A.1 Reactor Core Isolation Cooling System (RCIC) Actuation Instrumentation Minimum Operable Channels In the markup, Technical Specification Table 3.3.5-1, Reactor Core Isolation Cooling System Actuation Instrumentation, pages 3/4 3-45 and 3/4 3-46, specified the minimum operable channels per trip system as -
"2" and referred to " Action 50."
Action 50 stated "with the number of operable channels less then required by the Minimum Operable Channels per Trip System requirement:
For one trip system, place the inoperable channel in the tripped a.
condition within one hour or declare the RCIC system inoperable.
b.
For both trip systems, declare the RCIC system inoperable."
The initiation logic of RCIC at Grand Gulf Unit 1 is arranged as one trip system with four water level signals feeding a one-out-of-two-twice logic. The technical specification requirement of 2 minimum operable channels per trip system would not result in RCIC initiation
'unless the correct two channels are operable.
The minimum operable channels per trip system should have been four. Also, the Action Statement was intended for a 2-trip system design instead of the one-trip system design at Grand Gulf.
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Appendix.
Summary of Licensee Response The licensee admits, in part, that the initiation logic of the RCIC at Grand Gulf Unit 1 is for a one-trip system design and that the action statement was incorrectly written for a two-trip system design. However, MP&L does dispute the NRC staff's claim that this violation was material. The licensee's i
position is:
(1) the plant design did not change, (2) the minimum number of required operable channels did not change, (3) the remedial actions required in the event of channel inoperability did not change, and (4) RCIC initiation would have occurred as intended under the December 15, 1980 technical specification or the revised version of October l'1,1983.
The licensee points out that the definition of " trip system" for the RCIC system changed. The licensee states that the trip system information provided for the RCIC system in the December 15,19:' markup was correct for the definition of " trip system" in use at that time.
However, when NRC requested
)
information regarding the definition of a trip system in September 1983, the licensee clarified and standardized the boundaries of each trip system within the reactor protection system, thus redefining the boundary of the RCIC trip i
system.
During the Surveillance Review Program conducted by the licensee in 1983,.
several instrumentation questions concerning logic arrangements were raised by the NRC staff. As discussed in a letter from L.F. Dale (MP&L) to H.R. Denton (NRC) dated September 12, 1983, the licensee committed to define the terms
" channel," " trip system," and " trip function" as used in the technical specifications (TS) to resolve the NRC staff concerns.
In developing its response, the licensee standardized the definition of " trip system" so that one definition would apply to all TS instrumentation.
This resulted in redefining the RCIC initiation logic from what had been a two-trip system configuration consisting of four channels to a one-trip system configuration consisting of four channels. By letter dated October 11, 1983 from J. P. McGaughey (MP&L) to H.R. Denton (NRC), the licensee proposed TS for the RCIC actuation instrumentation consistent with this new definition.
In summary,.the licensee stated that the TS change for the RCIC actuation instrumentation proposed on October 11, 1983 was made to provide consistency with the new definition of trip system. The use of either the definition that existed when the December 15, 1980 TS markup was submitted, or the revised definition in effect at the time of the October 11, 1983 proposal to change TS, provides the same level of protection.
Appendix NRC Evaluation In reviewing the licensee's April 19, 1985 response to the Notice of Violation, the NRC staff has found significant new information, some of which is inconsistent with information provided in earlier submittals. Specifically, the September 12, 1983 submittal from the licensee stated:
2.
Concern: The NRC questioned whether or not single failure criterion was considered in determining the number of channels required to be operable for all cases in the Technical Specifications.
Response
The present Technical Specifications list Minimum Number of Channels per Trip System or Minimum Number of Channels per Trip Function for various instrumentation. As a part of the review effort required to provide the definitions of " channel," " trip system," and
" trip function," MP&L will provide verification that the single l
failure criterion has been properly applied where applicable....
Of the approximately 130 items reviewed, only two items were identified as potential problems with respect to the concerns identified by ICSB. As discussed below, only one of these necessitates a Technical Specification change.
a)
Table 3.3.5-1 lists the Octuation instrumentation for the Reactor Core Isolation Coeling (RCIC) System.
Presently the table requires two minimum operable channels per trip system for Reactor Vessel Water Level - Low Low, Level 2.
RCIC initiates on one-out-of-two-twice low level logic. To ensure a success path for initiation, all four low level channels are required operable. The RCIC system is not considered a safety related system, and consequently, single failure criteria is not requi red. However, because RCIC is redundant to HPCS for the safe shutdown function, and may be required under-certain abnormal conditions (FSAR Section 7.4.1.1.1.1), MP&L will submit necessary Technical Specification changes within thirty days of a
this letter. The proposed change will consist primarily of increasing the number of minimum operable channels to four.
(Emphasis added) b)
MP&L will evaluate increasing the minimum operable channels for actuation instrumentation from the present one per trip system to two per trip system.
This evaluation is being pursued not as a safety concern but to enhance system reliability since single failure criteria is met with the present Technical Specification requirements.
The October 11, 1983 submittal stated:
RCIC initiates on low reactor water level, Level 2.
The initiation logic is arranged as one trip system with four water level signals
D-Appendix feeding a one-out-of-two-twice logic.
The present requirement of two minimum OPERABLE channels per trip system does not ensure a success path for RCIC initiation unless the correct two channels are operable.
To insure that the success path is always maintained, the minimum OPERABLE channels per trip system for Reactor Vessel Water Level-Low Low, Level 2 should be increased to four which is the number of channels installed in the plant.
(Emphasis added)
Present ACTION 50 addresses two trip systems and does not reflect the "one-trip system" design. New ACTION 50 addresses the "one-trip system" design and matches the RCIC tr'p system design. New ACTION 50 allows up to two of four channels to be placed in the tripped condition before RCIC must be declared inoperable. This ACTION does not degrade system operability but is conservative since the trip system is closer to actuation with the channel (s) in the tripped condition.
This Technical Specification change is proposed as followup action to MP&L's discussions with Instrumentation and Control Systems Branch on this matter as committed to in AECM-83/0519, dated September 12, 1983.
Neither of these two submittals discusses the change in the definition of
" trip system" noted in the licensee's April 19, 1985 response to the Notice of Violation.
The April 19, 1985 response addresses the issues with the following:
The statement by MP&L in the October 11, 1983 submittal that the present requirement of two minimum OPERABLE channels per trip system does not ensure a success path for RCIC initiation unless the correct two channels are operable" applies to the situation that would have existed if the standardized definition of " trip system" had been changed but the corresponding operability requirements in the technical specifications had not been changed. However, since MP&L used the existing definition of
" trip system" accepted by GE until the technical specifications were revised, the situation described in the Notice of Violation (where RCIC could fail to operate) did not exist.
Based on a review of information provided in the April 19, 1985 subnittal,
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the NRC staff finds that MP&L's definition for RCIC trip system that existed at the time the December 15, 1980 markup was submitted and corresponding TS were consistent and provided an appropriate level of protection.
After further consideration and in light of the April 19, 1985 clarification that provided new information on the evolving definition discussed above in the Summary of Licensee Response, the staff finds that no material false statement as originally discussed in Example A.1 was made in the December 15, 1980 submittal.
It should be noted that if the licensee's December 15, 1980 and April 19, 1985 statements relative to RCIC logic are accurate, then the September 12, 1983 and October 11, 1983 statements are false. Accordingly, the staff could reissue this violation citing this fact. However, the staff has concluded that no additional benefit would be gained by doing so. Therefore, Example A.1 is withdrawn.
I I
s Appendix Restatement of Violation A (cont'd)
A.2 Core Alteration Requirements In the markup, Technical Specification 3.9.1 stated:
"The reattor mode switch shall be OPERABLE and locked in the Shutdown or Refuel position. When the reactor mode switch is locked in the Refuel position:
a.
A control rod shall not be inserted or withdrawn unless the Refuel position one-rod-out interlock is OPERABLE.
b.
CORE ALTEPATIONS shall not be performed using equipment associated with a Refuel position interlock unless the following associated Refuel position interlocks are OPERABLE for such equipment.
1.
All rods in.
2.
Refuel platform position.
3.
Refuel platform hoists fuel-loaded.
4.
Fuel grapple position.
5.
Source range monitor countrate."
However, Grand Gulf does not have a fuel grapple position interlock.
Summary of Licensee Response The licensee admits the violation. Although the Grand Gulf facility does not have a fuel grapple position interlock, one was listed in Technical Specification 3.9.1.b.4.
The licensee states that the use of BWR-6 model TS that reflected an alternate design approach contributed to this error.
The licensee references a March 24, 1983 letter from J.P. McGaughey (MP&L) to H.R. Denton (NRC) in support of this argument. The referenced letter provides l
the following discussion:
The fuel grapple not full up interlock has been deleted from the GGNS design. This interlock was a backup interlock to assure that no single failure could permit the refueling p.latform to be positioned over the core, a fuel assembly lifted, and a control rod to be inadvertently withdrawn. The need for this interlock was eliminated when redundant circuits were installed to sense the positioning of the refueling platform over the core and when the refueling platform main hoist is loaded with i
fuel. Since the fuel grapple position interlock has been deleted.in the GGNS Design, Technical Specification 3.9.1.b.4 should be deleted.
i In summary, the licensee adds that it believes the violation was minor and therefore does not warrant a civil penalty. The licensee requests that the j
civil penalty either be withdrawn or mitigated because of the lack of significance of the violation.
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Appendix NRC Evaluation With regard to the licensee's assertion that the model TS contributed to the f
error, the NRC expects licensees to review model TS for the licensee's design configuration. As evidenced by its markup of the model Technical Specification 3.9.1, the licensee did attempt to customize this specification for Grand Gulf.
However, the markup was incomplete and resulted in a TS that was inconsistent with the as-built design. The NRC staff does not believe that the error was made intentionally, rather it appears to be a result of f r attention to detail.
The licensee argues that even though this violation is adnitted, it does not warrant the severity level nor the civil penalty assessed by the NRC. When considering the licensee's explanation of April 19, 1985, the staff agrees that Example A.2, when considered by itself, was of minor safety significance.
Because the staff is withdrawing the other examples in Violation A, the violation is recategorized as a Severity Level IV violation with no divil penalty.
Restatement of Violation A (cont'd)
A.3 Discharge Line Pressures for ECCS Pumps
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In the markup, Technical Specification 4.5.1.b ECCS surveillance requirements for operability stated the required flow and discharge pressure to be:
I a.
LPCS pump develops a flow of at least 7115 gpm against a test line l
pressure greater than or equal to 128 psid.
j b.
LPCI pump develops a flow of at least 7450 gpm against a test line pressure greater than or equal to 111 psid.
4 c.
HPCS pump develops a flow of at least 7115 gpm against a test line pressure greater than or equal to 200 psid.
These technical specification discharge line pressure requirements for operability were subsequently changed to read respectively 290 psid,125 psid, and 445 psid. _The original discharge pressure requirements for operability were not consistent with the assumptions in the safety analysis.
3 Summary of Licensee Response i
The licensee admits that the original discharge pressure requirements for operability were not consistent with the assumptions presented in the safety analysis. However, the licensee adds that there were extenuating i
circumstances. Specifically, the basis for ECCS pump discharge pressures set forth in,the TS Surveillance Requirements was ambiguous and subject to t
interpretation. When the initial license was issued in June 1982, the basis j
for the ECCS pump discharge pressure values used by MP&L was system design 4
values. ASME Section XI testing was used as the basis for the technical y
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Appendix specification change submitted on June 19, 1983.
This basis was approved by the NRC in its safety evaluation attached to Amendment 9 to license NPF-29.
During the MP&L Technical Specification Review Program in 1984, the licensee noted that the ASME Section XI pump discharge test values did not satisfy worst-case safety analysis assumptions for ECCS flow and discharge pressure and, therefore, it proposed that safety analysis values be used instead. The April 18,1984 Order to NPF-29, provided final resolution.
In summary, the licensee states that the ambiguity as to the correct basis for this surveillance requirement resulted in a TS that was inconsistent with the assumptions of the safety analysis.
Based on the cause of the violation, the licensee requests mitigation of the proposed civil penalty.
NRC Evaluation The licensee argues that extenuating circumstances caused the discrepancy between the TS and the assumptions of the safety analysis. Based on a review of the information provided in the licensee's April 19, 1985 response, the NRC staff concludes that the licensee's statement was not false when considered in the circumstances within which it was made.
At the time that the December 15, 1980 markup was submitted, TS 4.5.1.b (ECCS flow / pressure tests) was relatively new.
The TS requirement was developed to supplement the inservice test requirements of 10 CFR 50.55a(g) and Section XI of the ASME Boiler and Pressure Vessel Code. These supplemental TS requirements were deemed necessary by the NRC staff when (in the later 1970s) detailed surveillance requirements were removed from the Standard Technical Specifications and a reference to Scaion XI of the Code was substituted in their place. The acceptance criteria of the Code permits an initial base-line pump output acceptance criterion and a certain amount of pump degradation over the service life. FSAR requirements are not factored into the code. Therefore, it is the responsibility of the licensee to ensure that the pump does not degrade below FSAR assumed pump parameters. The purpose of TS 4.5.1.b and other similar specifications is to periodically validate the assumptions of the safety analyses.
At the time that the h w
.5, 1980 markup was submitted, this basis for TS 4.5.1.b was not cleat 1y k u ited in any regulatory guidance.
Rather, the NRC staff relied on informal discussions between the NRC TS coordinator, NRC technical reviewer, and the applicant's staff to convey the basis.
It is not evident that this message was ever conveyed by the NRC staff. Accordingly, the staff' finds that this lack of a documented basis contributed to the error.
It should be noted that the December 15, 1980 values represented design differential pressures between the reactor pressure vessel and the suction water source for each pump and did not reflect the Code testing methods and acceptance criteria.
ASME Section XI testing requires verification of pump differential (i.e. total developed head) and flow at one point to be determined from preservice/ inservice reference data.
By letter dated June 29, 1983 from L. F. Dale (MP&L) to H. R. Denton (NRC), the licensee proposed a change to the TS to correct recognized errors and provide consistency with the Code requirements.
The justification accompanying the proposed change focused on Y
B l Appendix the Code testing methods and acceptance criteria.
In summarizing the justification for the proposed change the licensee stated: "This change does not introduce a significant reduction in a margin of safety...." Subsequent events indicate that the licensee did not thoroughly evaluate the margin of safety at the time that this statement was made.
It is' apparent that the NRC staff also failed to recognize the safety significance of the proposed TS change.
Via Amendment 9 to the TS the proposed change and supporting justifications were accepted by the NRC staff.
In its discussion of the change provided with the approved TS amendment, the NRC staff classified the change as administrative and noted that TS 4.5.1.b reflects test line pressure requirements to conform to ASME Section XI.
In 1984, during the MP&L Technical Specification Review Program, the licensee determined that the acceptance criteria of Surveillance Requirement 4.5.1.b could have allowed operation in a limited range of flow for one of the ECCS pumps that would not have met the safety analysis requirements. A proposed TS change to correct this error was submitted on April 10, 1984 and approved on April 18, 1984.
Based on its review of the information provided in the April 19, 1985 responses, the NRC staff finds that there were extenuating circumstances that contributed to the errors. The lack of NRC documented guidance to address a relatively new TS during TS development in December 1980 contributed to the first error.
Further, the incomplete guidance provided to the licensee during the review of Amendment 9 to the TS, as documented in the NRC's discussion accompanying the change, clearly contributed to the second error and lends additional credibility to the licensee's assertions regarding the ambiguity in the basis for TS 4.5.1.b in December 1980.
In summary, based on a reevaluation of the issue and a review of the clarifying information provided in the licensee's April 19, 1985 response, the statement described in this violation does not constitute a false statement and Example A.3 is withdrawn.
Restatement of Violation A (cont'd)
A.4 Class IE Distribution System In the markup, Technical Specification 4.8.1.1.1, Electrical Power Systems Surveillance Requirements, stated:
"Each of the above required independent circuits between the offsite transmission network and the onsite Class IE distribution system shall be:
b.
Demonstrated OPERABLE at least once per 18 months during shutdown by transferring manually and automatically, unit power supply from the normal circuit to the alternate circuit."
i l
. D:
1-Appendix 9-Grand Gulf Station did not have the. automatic transfer feature for offsite to onsite AC power sources.
I The false statement was material because if the NRC had known of the errors in the technical specifications, the NRC would not have issued the license with erroneous technical specifications.
I Summary of Licensee Response The licensee does not admit this violation and requests that the proposed violation and related civil penalty be withdrawn. The licensee's position is based on the fact that the December 15,'1980 markup of the Technical Specifications was correct at the time of its submittal. The Grand Gulf plant design was changed in October 1981 to delete the automatic transfer feature from the plant after the December 15, 1980 submittal.
ThelicenseepointsoutthatastatementcontainedintheNoticeofhiolation, l
" Grand Gulf did not have the automatic transfer feature for offsite to onsite ac power sources" is incorrect. The Grand Gulf design has always included this feature. The Notice of Violation should have addressed the design change that i
removed the automatic transfer from one offsite to another offsite (normal to
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alternate) power source at the ESF busses.
NRC Evaluation Even though the TS statement was correct when first submitted, the statement became false when modifications were made and the sta.ff was not promptly-informed of the changes. The staff used this erroneous information as a part of the NRC's review of the Grand Gulf TS. However, the staff agrees with the licensee that the substance of this violation has already been. incorporated in Violation B.
Consequently, Example A.4 is withdrawn.
Summary of NRC Evaluation of Violation A The December 15, 1980 markup of the TS submittal did contain technical errors.
Because Example A.1 has been withdrawn, Example A.2's severity level'has been.
changed from a Severity Level III to a Severity Level IV violation, and Examples A.3 and A.4 have been withdrawn, the civil penalty previously.
associated with Violation A is no longer appropriate.
Violation B l
Restatement of Violation B.
Inaccurate Submittals to the NRC On June 26,-1981 a second markup of the-Standard Technical Specifications for General Electric Boiling Water Reactors (NUREG-123),' Revision 2, August 1979 was submitted. This submittal contained the same errors as were contained'in the December 15, 1980 submittal.. Additional 1
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i submittals and changes were transmitted in letters ~ dated December 31, 1981, January 12, February 25, March 23, April 5, 6., 7 and 30, May 26, June 1, 9 (two letters) and 10, 1982. On June 16, 1982, NRC issued a low power '(5".) license. to MP&L for Grand Gulf Unit I with appended technical specifications based upon the licensee's submittals.
The errors con-tained in Item A were reflected in the technical specifications issued with the license.
Each of the submittals was a separate opportunity to discover and correct the false submittal of December 15, 1980.. The. licensee's-failure to do so constitutes a material false statement by omission.
The statement was false because the licensee failed to correct its initial false submittal and to ensure that the technical specifications ultimately-'
i
,2 issued with its license reflected plant-specific design features. The j
statement was material because the NRC would not have issued the license i
with erroneous technical specifications.
l Summary of Licensee Response l
The licensee admits the violation and states that it was caused by a review
- process that was not sufficiently systematic or rigorous. The licensee acknowledges that the June 21, 1981 submittal was-esse'ntially an entire draft that provided an opportunity to correct erroneous information.
However, the.
licensee disagrees with the NRC staff's statement that each of the TS revisions requested subsequent to the June 26, 1981 submittal was a separate opportunity:
to discover and correct the. original false submittal. As -a basis for the disagreement, the licensee states that the December 1980 and June-1981' submittals were complete " mark-up" revisions of TS whereas subsequent-submittals only involved isolated sections and did not' necessitate a complete TS review. Therefore, the licensee claims that those-submittals should not be categorized as " missed opportunities." Thus, the licensee requests that the t
' civil penalty be mitigated.
NRC Evaluation The NRC staff maintains that the licensee has a duty to ensure that submittals made to the NRC are accurate.
The licensee states that each of the submittals
-subsequent to the December-15, 1980 and June 26,'1981 submittals were not separate opportunities to discover and correct false information. 'As the NRC' has stated, three of the four examples presented as al basis for Violation A have been withdrawn. However, it should be noted that Violation A only cited some of the examples of erroneous information submitted-to the NRC.
Example-
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A.1 was based on the licensee's own incorrect statements in September 12 and October 11,-1983 submittals to the NRC. The. proposed violation was-withdrawn-because of an incorrect basis cited for the' original violation. This-violation E
could have been_ reissued with a correct basis, although there would:be minimal i
. benefit in doing so. Example A.2 was of minor safety. significance and was1 recategorized as a Severity Level IV violation.. Example A.3 was withdrawn rr v
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Appendix after reviewing information contained in the licensee's April 19, 1985 1
i submittal. Example A.4 was withdrawn but still involved the licensee not notifying the staff of a modification to the facility. However, since the remaining example for Violation A (Example A.2) used as the basis for Violation B has been recategorized at a Severity Level IV and the staff does not propose to issue a new Notice with other examples, Violation B is also recategorized at a Severity Level IV and the civil penalty previously associated with Violation B is no longer appropriate.
Restatement of Violation i
Violation C C.
In a letter to Harold Denton dated June 14, 1983, Mississippi Power and Light Company provided additional proposed changes to the Grand. Gulf Technical Specifications which MP&L stated that the changes were
" intended, in general, to enhance clarity or provide consistency with the plant design and operation."
i Contrary to Section 186 of the Atomic Energy Act, this letter contained material false statements. The statements were false as shown below:
C.1 Load Shedding and Sequencing Panel With reference to Technical Specification 4.8.1.1.1.b the submittal 4
stated, "There is no automatic transfer from the normal to alternate i.
circuit since this bus search and automatic transfer feature was deleted 2
from the load shedding and sequencing (LSS) panel by a pre-operating license design change. Section 8.3 of the FSAR no longer contains a description of the bus search and automatic transfer feature of the LSS panel."
4 However, the LSS panel still performs a. search and automatic transfer.
function even though it is prohibited from auto-transfer hookup to another offsite power source by other features.
Summary of Licensee Response
~
The licensee admits that the load shedding and sequencing panel still contains the electronics which could perform a search-and-automatic-transfer function if 1
signal cables were reconnected. The licensee considers the " disabling" of this feature to be equivalent to the " deletion" of the feature because what was done prevented breaker closure and connection to other offsite sources.
The design change which disabled the bus search-and-automatic transfer. feature was performed by disconnecting signal cable connections.to ensure that the j
feature would not function.
This design change was made to ensure that i
breakers to other offsite power sources would not automatically close and supply power to the ESF buses. Both the Architect Engineer for Grand Gulf and the-licensee' consider this disabling type design change equivalent to deleting the. function.
Based on this reasoning, the licensee requests mitigation of the j
proposed civil. penalty.
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5 Appendix NRC Evaluation Based on its review of the information provided in the licensee's April 19, 1985 response, the NRC staff agrees that the disabling of the electronics did in fact delete the feature.
Because disabling was done through a formal design change process that included several stages of review and approval, drawing changes, and quality assurance controls and because the reconnection of 4
this feature would be subject to the same process, the NRC staff considers this a deletion.
In summary, based on the clarifying information provided in the licensee's April 19,1985 response, the statement described in this violation does not constitute a material false statement and Example C.1 is withdrawn.
Restatement of Violation C (cont'd)
C.2 Organization With reference to Technical Specification 6.5.2.2, the submittal stated that the Manager of Systems Nuclear Operations, Middle South Services, will be replaced by a qualified representative of System Nuclear Operations.
However, the organizational entity, Systems Nuclear Operations, did not exist.
The statements were material because an agency reviewer would not have made the requested changes to the technical specifications had the reviewer known that the bases for the changes were wrong.
Summary of Licensee Response The licensee admits this violation. On the basis that the error was in an administrative section of the TS, the licensee believes that the violation should be reclassified to a lesser severity level.
The licensee points out that en early 1983 reorganization occurred within Middle South Services that functionally made what had been called Systems Nuclear Operations a part of Nuclear Assurance.
In the licensee's opinion, this contributed to the error. Further, even though the June 14, 1983 submittal reflected an error in the title of an individual, the qualifications for the position still had to be met.
NRC Evaluation It is the licensee's responsibility to ensure that information contained in a TS change request is accurate.
In response to the licensee's arguments, the NRC staff reiterates that it considered the title change as it was proposed.
Although the change was denied by NRC for reasons not directly related to the' s
error, the fact remains that the proposal was in error.
(This change was subsequently corrected and approved in Amendment 9 to the TS.) However, based on a reevaluation of this issue, the staff believes that this example was of such minor significance when considered by itself that it should be-recategorized as a Severity Level IV violation with no civil penalty.
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O Appendix Summary of NRC Evaluation of Violation C Because Example C.1 has been withdrawn and Example C.2's severity level has been changed from a Severity Level III to a Severity Level IV, the civil penalty previously associated with Violation C is no longer appropriate.
Violation D Restatement of Violation D.
Instrument Channel Calibration In a letter to the Nuclear Regulatory Commission staff dated June 23, 1983, Mississippi Power and Light Company provided additional proposed changes to the Grand Gulf Technical Specifications which MP&L stated were
" intended, in general, to enhance clarity or provide consistency with the plant design and operation."
Contrary to Section 186 of the Atomic Energy Act, this letter contained a material false statement. The statement was false as shown below:
The submittal stated, with reference to Technical Specification Table 4.3./.5-1, that a note requiring channel calibration does not apply to the instruments used at Grand Gulf and should be deleted.
i However, the footnote fully applied to the Grand Gulf instruments.
l The false statement was material because the reviewer might have made the requested change to the technical specifications based upon incorrect information.
Summary of Licensee Response The licensee acknowledges that the June 23, 1983 proposed TS change included a statement that a note on TS Table 4.3.7.5-1 did not apply to the instruments being used at Grand Gulf and should be deleted. However, the licensee questions whether the June 23, 1983 submittal containing this wording was a
" false statement." This is based on the licensee's claim that this matter arose out of a technical misunderstanding with the NRC staff.
In the June 23, 1983 submittal the licensee proposed to calibrate the instruments using a sample gas that met the instrument vendor's recommendations.
Rather than having the details of the test in the TS, the licensee proposed to address them in surveillance procedures. The licensee believes that the technical basis that it had for deleting the footnote was valid. However, when the NRC disagreed with the deletion as a technical matter, the licensee agreed to adopt the position of the NRC for a two point calibration. This did not establish that the licensee's technical basis was invalid or. that any " false" information had been submitted.
O Appendix NRC Evaluation After further consideration of this violation, the staff agrees with the licensee that the June 23, 1983 letter merely contained a proposal to delete a calibration method from the TS.
In the disussion justifying the deletion, the licensee informeo the staff that calibration of the containment and drywell hydrogen concentration analyzers would be performed using a different sample gas concentration. The licensee's statement that the calibration requirements do not apply to the instruments being used at Grand Gulf could be construed as a material false statement when taken out of context. However, when viewed in context, it merely states that it was the licensee's position that a particular method of calibration should not apply to instruments at the facility. This was based on manufacturer's recommendations and the licensee's preference to address the details of the calibration in plant procedures rather than in the TS. Therefore, the statement described in this violation does not constitute a material false statement and Violation D is withdrawn.
Violation E Restatement of Violation E.
Circuit Breakers In a letter to Harold Denton dated August 5,1984, Mississippi Power and Light certified that the Grand Gulf Technical Specifications transmitted to the NRC up to and including Amendment 13 accurately reflected the plant, the FSAR and supporting documents and the SER in all material respects.
Contrary to Section 186 of the Atomic Energy Act of 1954, as amended, this statement was a material false statement. The statement was false because the technical specifications did not reflect existing plant-specific design features as shown below.
In a letter to the NRC dated August 14, 1984, MP&L requested additional changes to the Grand Gulf Technical Specifications to add circuit breakers to the list of those circuit breakers performing primary containment penetration conductor-overcurrent protection functions for which surveillance was required.
The false statement was material because the NRC believed the list of circuit breakers requiring surveillance was complete and might have issued the license with erroneous technical specifications, had the licensee not subsequently corrected the error.
Summary of Licensee Response The licensee acknowledges that additional changes adding circuit breakers to TS Table 3.8.4.1-1 were necessary to complete the table even though it
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had already certified the accuracy of the TS.
The licensee notes that the i
evolutionary process of determining the appropriate protection required contributed to the errors. The licensee believes the discrepancy was not significant because a second level of circuit protection existed. For this reason and because the licensee discovered and promptly corrected the discrepancy, the licensee requests that the proposed civil penalty be mitigated.
1 Appendix NRC Evaluation The licensee certified to the staff in a [[letter::AECM-84-0394, Forwards Supplemental Info Re Qualification of Safety Relief Valves (Srv),Seitz Solenoid Valves.Interim Operation Justified Until SRV Solenoid Valves Fully Qualified,Per NUREG-0588 & 10CFR50.49|letter dated August 5,1984]] that the TS submitted to the staff accurately reflected the plant, the FSAR, and the SER. Subsequent to that certification, as a result of an update of a Unit 2 design drawing to incorporate Unit I changes, it was determined that a discrepancy existed between the FSAR and the Unit 1 plant design.
Design modifications were initiated and associated TS proposed.
Based on a review of the clarifying information provided in the licensee's April 19, 1985 response it is apparent that a revised licensing basis in 1978 (upgrade from RG 1.63, Rev. O to RG 1.63, Revision 1) did contribute to the discrepancy between the as-built plant, the FSAR and the TS. The licensee admits and the NRC staff agrees that a proper review of the circuit design for adequacy of overcurrent protection was not performed before the August 5,1984 certification. However, since the licensee continued its review of the TS, discovered the discrepancies, and corrected them within nine days after providing the inaccurate information in the [[letter::AECM-84-0394, Forwards Supplemental Info Re Qualification of Safety Relief Valves (Srv),Seitz Solenoid Valves.Interim Operation Justified Until SRV Solenoid Valves Fully Qualified,Per NUREG-0588 & 10CFR50.49|August 5,1984 letter]] and because of the mistakes made in the other citations, the proposed civil penalty will be withdrawn.
Conclusion After carefully considering the licensee's response as discussed above, the determination has been made that 1) examples A.1, A.3, and A.4 of Violation A are withdrawn; 2) the Severity Level for the remaining example, A.2, is changed from a Severity Level III to a Severity Level IV with no civil penalty;
- 3) Violation B and example C.2 of Violation C are changed from Severity Level III to Severity Level IV with no civil penalties; 4) example C.1 of Violation C and Violation D are withdrawn; and 5) the civil penalty for Violation E is withdrawn.
Appendix NRC Evaluation Based on its review of the information provided in the licensee's April 19, 1985 response, the NRC staff agrees that the disabling of the electronics did in fact delete the feature.
Because disabling was done through a formal design change process that included several stages of review and approval, drawing changes, and quality assurance controls and because the reconnection of this feature would be subject to the same process, the NRC staff considers this a deletion.
In summary, based on the clarifying information provided in the licensee's April 19, 1985 response, the statement described in this violation does not constitute a material false statement and Example C.1 is withdrawn.
Restatement of Violation C (cont'd)
C.2 Organization With reference to Technical Specification 6.5.2.2, the submittal stated that the Manager of Systems Nuclear Operations, Middle South Services, will be replaced by a qualified representative of System Nuclear Operations.
However, the organizational entity, Systems Nuclear Operations, did not exist.
The statements were material because an agency reviewer would not have made the requested changes to the technical specifications had the reviewer known that the bases for the changes were wrong.
Summary of Licensee Response The licensee admits this violation. On the basis that the error was in an administrative section of the TS, the licensee believes that the violation should be reclassified to a lesser severity level.
The licensee points out that an early 1983 reorganization occurred within Middle South Services that functionally made what had been called Systems Nuclear Operations a part of Nuclear Assurance.
In the licensee's opinion, this contributed to the error.
Further, even though the June 14, 1983 submittal reflected an error in the title of an individual, the qualifications for the position still had to be met.
NRC Evaluation It is the licensee's responsibility to ensure that information contained in a TS change request is accurate.
In response to the licensee's arguments, the NRC staff reiterates that it considered the title change as it was proposed.
Although the change was denied by NRC for reasons not directly related to-the
-error, the fact remains that the proposal was in error.
(This change was subsequently corrected and approved in Amendment 9 to the TS.) However, based on a reevaluation of this issue, the staff believes that this example was of such minor significance when considered by itself that it should be recategorized as a Severity Level IV violation with no civil penalty.
Appendix Summary of NRC Evaluation of Violation C Because Example C.1 has been withdrawn and Example C.2's severity level has been changed from a Severity Level III to a Severity Level IV, the civil penalty previously associated with Violation C is no longer appropriate.
Violation D Restatement of Violation D.
Instrument Channel Calibration In a letter to the Nuclear Regulatory Commission staff dated June 23, 1983, Mississippi Power and Light Company provided additional proposed changes to the Grand Gulf Technical Specifications which MP&L stated were
" intended, in general, to enhance clarity or provide consistency,with the plant design and operation."
Contrary to Section 186 of the Atomic Energy Act, this letter contained a material false statement. The statement was false as shown below:
The submittal stated, with reference to Technical Specification Table 4.3.7.5-1, that a note requiring channel calibration does not apply to the instruments used at Grand Gulf and should be deleted.
However, the footnote fully applied to the Grand Gulf instruments.
The false statement was material because the reviewer might have made the requested change to the technical specifications based upon incorrect information.
Summary of Licensee Response The licensee acknowledges that the June 23, 1983 proposed TS change included a statement that a note on TS Table 4.3.7.5-1 did not apply to the instruments being used at Grand Gulf and should be deleted. However, the licensee questions whether the June 23, 1983 submittal containing this wording was a
" false statement." This is based on the licensee's claim that this matter arose out of a technical misunderstanding with the NRC staff.
In the June 23, 1983 submittal the licensee proposed to calibrate the instruments using a sample gas that met the instrument vendor's recommendations. Rather than having the details of the test in the TS, the licensee proposed to address them in surveillance procedures. The licensee believes that the technical basis that it had for deleting the footncte was valid. However, when the NRC disagreed with the deletion as a technical matter, the licensee agreed to adopt the position of the NRC for a two point calibration.
This did not establish that the licensee's technical basis was invalid or that any " false" information had been submitted.
a Appendix NRC Evaluation After further consideration of this violation, the staff agrees with the licensee that the June 23, 1983 letter merely contained a proposal to delete a calibration method from the TS.
In the disussion justifying the deletion, the licensee informed the staff that calibration of the containment and drywell hydrogen concentration analyzers would be performed using a different sample gas concentration. The licensee's statement that the calibration requirements do not apply to the instruments being used at Grand Gulf could be construed as a material false statement when taken out of context. However, when viewed in context, it merely states that it was the licensee's position that a particular t
method of calibration should not apply to instruments at the facility. This was based on manufacturer's recommendations and the licensee's preference to address the details of the calibration in plant procedures rather than in the TS. Therefore, the statement described in this violation does not constitute a material false statement and. Violation D is withdrawn.
Violation E Restatement of Violation E.
Circuit Breakers In a letter to Harold Denton dated August 5,1984, Mississippi Power and Light certified that the Grand Gulf Technical Specifications transmitted to the NRC up to and including Amendment 13 accurately reflected the plant, the FSAR and supporting documents and the SER in all material respects.
Contrary to Section 186 of the Atomic Energy Act of 1954, as amended, this statement was a material false statement. The statement was false because the technical specifications did not reflect existing plant-specific design features as shown below.
In a letter to the NRC dated August 14, 1984, MP&L requested additional changes to the Grand Gulf Technical Specifications to add circuit breakers to the list of those circuit breakers performing primary containment penetration conductor overcurrent protection functions for which surveillance was required.
The false statement was material because the NRC believed the list of circuit breakers requiring surveillance was complete and might have issued the license with erroneous technical specifications, had the licensee not subsequently corrected the error.
Summary of Licensee Response The licensee acknowledges that additional changes adding circuit breakers to TS Table 3.8.4.1-1 were necessary to complete the table even though it had already certified the accuracy of the TS. The licensee notes that the evolutionary process of determining the appropriate protection required contributed to the errors. The licensee believes the discrepancy was not significant because a second level of circuit protection existed. For this 4
reason and because the licensee discovered and promptly corrected the i
discrepancy, the licensee requests that the proposed civil penalty be mitigated.
Appendix NRC Evaluation The licensee certified to the staff in a [[letter::AECM-84-0394, Forwards Supplemental Info Re Qualification of Safety Relief Valves (Srv),Seitz Solenoid Valves.Interim Operation Justified Until SRV Solenoid Valves Fully Qualified,Per NUREG-0588 & 10CFR50.49|letter dated August 5,1984]] that the TS submitted to the staff accurately reflected the plant, the FSAR, and the SER. Subsequent to that certification, as a result of an update of a Unit 2 design drawing to incorporate Unit 1 changes, it was determined that a discrepancy existed between the FSAR and the Unit 1 plant design.
Design modifications were initiated and associated TS proposed.
Based on a review of the clarifying information provided in the licensee's April 19,1985 respense it is apparent that a revised licensing basis in 1978 (upgrade from RG 1.63, Rev. O to RG 1.63, Revision 1) did contribute to the discrepancy between the as-built plant, the FSAR and the TS. The licensee admits and the NRC staff agrees that a proper review of the circuit design for ndequacy of overcurrent protection was not performed before the August 5,1984 sertification. However, since the licensee continued its review of the TS, discovered the discrepancies, and corrected them within nine days after providing the inaccurate information in the [[letter::AECM-84-0394, Forwards Supplemental Info Re Qualification of Safety Relief Valves (Srv),Seitz Solenoid Valves.Interim Operation Justified Until SRV Solenoid Valves Fully Qualified,Per NUREG-0588 & 10CFR50.49|August 5,1984 letter]] and because of the mistakes made in the other citations, the proposed civil penalty will be withdrawn.
Conclusion After carefully considering the licensee's response as discussed above, the i
determination has been made that 1) examples A.1, A.3, and A.4 of Violation A are withdrawn; 2) the Severity Level for the remaining example, A.2, is changed from a Severity Level III to a Severity Level IV with no civil penalty;
- 3) Violation B and example C.2 of Violation C are changed from Severity Level III to Severity Level IV with no civil penalties; 4) example C.1 of Violation C and Violation D are withdrawn; and 5) the civil penalty for Violation E is withdrawn.
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